AP1000 Nuclear Power Plant Research Articles

Seismic fragility analysis of shield building considering strength ratio of mainshock and aftershocks

The shield building of the AP1000 nuclear power plant serves as a crucial protective barrier against radioactive substances. However, past research indicates that structures are susceptible to experiencing aftershocks, which may lead to unforeseeable damage and potential radioactive material leakage. To address this issue, a finite element model of the shield building was established with the damage indexes of the tensile and compressive damage selected for further model analysis. According to the fundamental theory of reliability, the traditional incremental dynamic analysis method was used to analyze the seismic fragility of the shield building by inputting mainshock and aftershock sequences with three strength ratios. The results indicate that the seismic fragility of shield building may be underestimated without considering the influence of aftershocks and the damage state presents an upward tendency as the strength ratio increases. However, the cumulative damage caused by aftershocks is unlikely to exceed the initial damage induced by the corresponding mainshock. Overall, the aggravation of the compressive damage is less pronounced than the increase of the tensile damage as the strength ratio increases.

An improved lump mass stick model of a nuclear power plant based on the Kriging surrogate model

This paper focuses on an improved lump mass stick model of the AP1000 nuclear power plant (NPP). The conventional method is used to calculate the parameters of the lump mass stick (LMS) model of the AP1000 NPP, which has great differences between the LMS model and the finite element (FEM) model in terms of frequency and vibration mode. Therefore, with optimizing the mass condensation size of a shear wall without a floor as a goal, the improved lump mass stick (I-LMS) model based on the Kriging surrogate model is established. The results illustrate that: (i) compared with the LMS model, the relative error of the I-LMS model and the FEM model, the maximum error of the frequency is reduced from 23.04% to 6.57% with nine order frequencies; (ii) the frequencies of the Kriging surrogate model and the I-LMS model are very close, with the largest error being only 1.45%, indicating that the Kriging surrogate model has obtained better correction accuracy; (iii) the modal assurance criterion (MAC) value of the I-LMS model for the same vibration mode is increased from 0.75 to 0.95 to 0.88–0.96, and the MAC values of different vibration modes are closer to 0; and (iv) comparison of the structural dynamic response under basic design earthquake to the FEM model shows that the I-LMS model has good calculation accuracy and obvious advantages in calculation efficiency.

Investigations on seismic performance of nuclear power plants equipped with an optimal BIS-TMDI considering FSI effects

This paper introduces a base isolation system-tuned mass damper inerter (BIS-TMDI) hybrid system to the AP1000 nuclear power plant (NPP), which reduces seismic damage potential of the NPP structure. The effects of fluid-structure interaction (FSI) caused by the passive containment cooling system water storage tank (PCCWST) on NPP's seismic performance are investigated. The FSI of water tank theoretical model is considered based on the Housner's model, and a series of time history analyses are performed to prove the rationality of the proposed model. Three single-objective optimization strategies are employed to minimize the relative displacement variance and absolute acceleration variance of the upper structure, as well as the filtered energy index (FEI). Furthermore, a multi-objective optimization strategy considering all these three indexes is proposed to obtain optimal parameters of vibration control. The influence of vibration control strategies on the relative deformation and acceleration of the upper structure is explored with various water level ratios. The analytical results indicate that the proposed BIS-TMDI strategy has significantly reduced the NPP structure's seismic response. The effectiveness of the vibration control strategy is influenced by the water level ratio, emphasizing the significance of designing an appropriate water level ratio to reduce NPP structure's seismic response.

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Safety Analysis in AP-1000 Nuclear Power Plant Against Primary Loop Coolant Pump and Fuel Failures Authors

Calculation and analysis of fast neutron fluence of pressure vessel in AP1000 reactor

The structural integrity of the reactor pressure vessel plays an important role in the safe operation of AP1000 nuclear power plant. Radiation damage of fast neutrons is the main factor leading to pressure vessel material deterioration. Therefore, it is necessary to accurately calculate the fast neutron fluence in the pressure vessel. Discrete ordmates method radiation shielding analysis code cosSN is used to calculate fast neutron fluence hi the pressure vessel and the calculation result is compare to measurement of actual AP1000 power plant. Bondarenko method is adopted to calculate multi-group cross section hi resonance energy region, and Bell-Hansen-Sandmeier(BHS) transport correct method is used to deal with anisotropic scattering effect. To improve computational efficiency, cosSN adopts the diffuse comprehensive acceleration and the multi-grid technique to accelerate source iteration, and adopts parallel computing to solve the storage limitations in large-scale calculations. For fast neutron fluence calculation, the calculation uncertainty conies from input parameters and approximation of theoretical model. The results hidicate that the calculated specific activity is conform to the actual measured value of the power plant. The calculation uncertainty is ± 13.28%. which meets the requirements of China domestic energy industry standards NB/T 20576-2019 and NRC RG 1.190.

Stochastic dynamic and reliability analysis of AP1000 nuclear power plants via DPIM subjected to mainshock-aftershock sequences

Earthquakes usually consist of a mainshock and a series of aftershocks, both of which are random. The mainshock can damage structures, and the subsequent aftershocks may cause further damage. However, no studies have considered the effects of stochastic seismic sequences on nuclear power plants (NPPs). We propose a new framework to analyze the stochastic dynamic response and dynamic reliability of AP1000 NPPs. First, stochastic mainshock-aftershock sequences are generated by combining a physical random function model, the narrowband harmonic group superposition method and the copula function. Then, the effects of aftershocks on NPPs are highlighted using acceleration and displacement as response indices and the tensile damage ratio (TDR) and plastic strain as damage indicators. Finally, probabilistic information and reliability of NPPs are obtained based on the direct probability integration method (DPIM) and the absorbing condition method (ACM). The results show that the dynamic response is larger for seismic sequences than a single mainshock, and seismic sequences cause more damage to the NPPs. When the peak ground acceleration (PGA) increases, the effect of the aftershock becomes more pronounced, and the response of the NPP becomes more random. Moreover, aftershocks reduce the reliability of NPPs, and the degree of reduction is related to the thresholds.

Shaking table test study of nuclear power plant model considering soil-structure interaction effect

The influence analysis of site condition is fundamental to the seismic design of nuclear power plants. In this paper, a series of shaking table tests were performed to investigate the Soil-Structure Interaction (SSI) effect on the seismic response of nuclear power structures. An AP1000 nuclear power plant was taken as the objective structure, lead rubber bearings (LRBs) were used as the isolators and Shanghai silty clay was selected as the model soil. Test results show that (1) The soil has different effects on the dynamic characteristics of isolated and non-isolated model structures. (2) During the shaking table test, the foundation edge is subject to the highest soil pressure, and a contact surface separation between the foundation and the soil begins to occur under the seismic-margin earthquake (SME) seismic input level. (3) Both for the rigid and soil foundation, seismic isolation bearings can effectively reduce the acceleration response of the superstructure. (4) For the non-isolated model, the SSI effect significantly amplifies the seismic response at the bottom floor of the structure in the low frequency range, and an obvious rocking behavior occurs at the operating basis earthquake (OBE) seismic input level. (5) For the isolated model, the SSI effect reduces the hysteresis capacity of the seismic isolation system and amplifies the seismic response at the bottom floor, especially in the vertical direction. The site condition and the use of an isolation system have a significant impact on the seismic response of a nuclear power structure. Therefore, in the design and safety assessment of nuclear power plants, the influences of soil characteristics and isolation equipment in the soil-structure interaction system should be fully considered.

Fragility analysis of nuclear power plant structure under real and spectrum-compatible seismic waves considering soil-structure interaction effect

Both the real and spectrum-compatible seismic waves are widely used as input loads in the seismic assessment of critical structures in engineering applications, which are also allowed by international codes. However, the effect of the real seismic waves and spectrum-compatible seismic waves generated by different methods on the dynamic responses of structures has not been thoroughly emphasized. This study focused on quantitatively comparing the dynamic responses and fragility of the AP1000 nuclear power plant under real and spectrum-compatible seismic waves. Real and spectrum-compatible seismic waves generated by the modification and synthetic methods were used to perform the time history analysis. The effects of three input motion types on a high-accuracy soil–structure interaction system model were investigated based on various dynamic response indexes. The fragility assessment and maximum tension strain contours of the AP1000 nuclear power plant subjected to the three types of seismic waves were also presented to evaluate the influence of the input motions. The results revealed that spectrum-compatible seismic waves can cause larger dynamic responses than real seismic waves at the same peak ground acceleration level. Additionally, although some response indexes, such as strain energy, are sensitive to the input motions when the nuclear power plant is in a nonlinear state, the generation methods of spectrum-compatible seismic waves have insignificant effects on most dynamic response indexes, fragility assessment, and damaged zone locations.

Порівняльний метод кваліфікації систем безпеки ядерних енергоустановок з ВВЕР‑1000 та АР1000

A deterministic method of analyzing conditions for ensuring the safe operation of VVER-1000 and AP1000 nuclear power plants has been developed. The method is based on the comparison of the provision of the required safety functions by critical (minimum permissible) configurations of safety systems for accident management. During the analysis, the composition and information on the technical characteristics of the safety systems of the VVER-1000/320 and AP1000 reactor installations, which are designed to ensure the safety functions of feeding and cooling the reactor/PG, were considered. On the basis of the developed method, the relative criteria and safety conditions (for example — the criterion of safety — the maximum temperature of the heating element shell; the condition of safety is not to exceed the maximum design limit of damage to the twelfth (the temperature of the twelfth shell is 1,200 °C) for nuclear power plants with VVER-1000 and АР1000 reactors in terms of safety functions for emergency fueling and cooling of the reactor core and steam generators are determined. Preliminary estimates of the relative criteria and conditions for ensuring nuclear safety during accident management of natural circulation circuits of passive safety systems AP1000 obtained taking into account data on their composition and technical characteristics, the table showed that the required height of natural circulation circuits of passive safety systems should be several hundred meters. Based on the established relative criteria and safety conditions, the need for additional modernization of the АР1000 safety systems to successfully ensure the safety functions of emergency coolant/feedwater supply and core and steam generator cooling is preliminarily determined.

An improved time-domain approach for the spectra-compatible seismic motion generation considering intrinsic non-stationary features

The dynamic structural responses are sensitive to the time-frequency content of seismic waves, and seismic input motions in time-history analysis are usually required to be compatible with design response spectra according to nuclear codes. In order to generate spectra-compatible input motions while maintaining the intrinsic non-stationarity of seismic waves, an improved time-domain approach is proposed in this paper. To maintain the nonstationary characteristics of the given seismic waves, a new time-frequency envelope function is constructed using the Hilbert amplitude spectrum. Based on the intrinsic mode functions (IMFs) obtained from given seismic waves through variational mode decomposition, a new corrective time history is constructed to locally modify the given seismic waves. The proposed corrective time history and time-frequency envelope function are unique for each earthquake records as they are extracted from the given seismic waves. In addition, a dimension reduction iterative technique is presented herein to simultaneously superimpose corrective time histories of all the damping ratios at a specific frequency in the time domain according to optimal weights, which are found by the genetic algorithm (GA). Examples are presented to show the capability of the proposed approach in generating spectra-compatible time histories, especially in maintaining the nonstationary characteristics of seismic records. And numerical results reveal that the modified time histories generated by the proposed method can obtain similar dynamic behaviors of AP1000 nuclear power plant with the natural seismic records. Thus, the proposed method can be efficiently used in the design practices.

Impact response prediction and optimization of SC walls using machine learning algorithms

The dynamic impact response of steel-plate composite (SC) walls is a vital index to assess their safety. By optimizing the design parameters within the requirements of relevant specifications, the impact-induced deformation can be reduced to ensure the protection of personnel and equipment inside the structures. To process the complicated nonlinear relationship between the wall deformation and design parameters rapidly and accurately, prediction models were constructed through such machine learning algorithms as support vector regression (SVR), artificial neural network (ANN), and gaussian process regression (GPR) in this study. Then these models were further validated by the existing test results. It shows that the root mean square error (RMSE), R-Squared (R2), and the time to complete one calculation of the GPR model are 2.74, 0.9591, and 0.58 ms respectively, which has obvious advantages in terms of accuracy, stability, and efficiency among the three machine learning models. Based on the well-trained GPR model, the design of a certain SC wall in an AP1000 nuclear power plant was optimized via the genetic algorithm, with the maximum deformation and the cost of SC walls as the optimization targets. The case study shows that the optimal solution set can reduce the maximum deformation by 10.1 % while keeping the total cost constant, or reduce the total cost by 18.8 % while keeping the maximum deformation constant.

Innovation structure combining inter-story isolation with passive cooling effect for AP1000 nuclear power plants

PurposeAn inter-story isolation structure (IIS) for AP1000 nuclear power plants (NPPs) is provided to resolve the conflict of seismic safety and the optimal location of air intakes.Design/methodology/approachThe effect of passive cooling system (PCS) is better with lower altitude of air intakes than that in the original design of AP1000 NPPs. Seismic performances of IIS NPPs, including the seismic responses, damping frequency bandwidth and seismic reduction robustness, are improved by combining the position of air intakes lower and the optimal design method.FindingsTheoretical analysis and numerical simulation are illustrated that the seismic reduction failure of IIS NPPs is the lowest probability of occurrence when PCS has highest working efficiency.Originality/valueThe IIS NPPs can transfer the contradiction between PCS work efficiency and seismic safety of NPPs to the mutual promotion of them.

Seismic fragility analysis of nuclear power plants considering structural parameter uncertainty

This paper proposes a new method for seismic fragility analysis of nuclear power plants (NPPs) considering the uncertainty sources of both ground motions and structural parameters. Based on the conventional seismic fragility model that only contains the uncertainty of ground motions, seismic fragility analysis conditioned on a specific seismic intensity is probabilistically reformulated by further incorporating the uncertainty of structural parameters. To improve the efficiency of the analysis, a point estimation-assisted incremental dynamic analysis was developed for evaluating the seismic fragility model including the uncertainty of structural parameter. The accuracy and efficiency of the proposed method were demonstrated by a single-degree-of-freedom system, where Monte Carlo simulation method is used for comparison. The seismic fragility based on the finite element model of an AP1000 NPP was analyzed using the proposed method. The results reveal that the safety of NPP structure would be overestimated without considering the uncertainty of structural parameter and the influence of the uncertainty of structural parameters on the seismic fragility of NPP presents an upward trend as the damage level increases. Furthermore, the sensitivity analysis of different structural parameters reveals that elastic module may have a dominant influence.

Seismic fragility analysis of nuclear power plants based on the substructure method

This paper focuses on seismic fragility analysis of the AP1000 Nuclear Power Plant (NPP) with a substructure (SUB) model. To improve the efficiency of the analysis and calculation, this study treats the steel vessel containment (SVC) and equivalent equipment as a substructure and connects them with the residual structure to form an SUB model. The SUB model is verified to determine its rationality. Based on the verified model, 160 working conditions are calculated, and an incremental dynamic analysis (IDA) is performed. The AP1000 NPP is divided into 11 zones and 6 layers to investigate the seismic fragility analysis in detail. The results illustrate that (i) the calculation time of the SUB model is reduced by 1/4 compared to that of the full finite element model of the AP1000 NPP (FULL model) under the premise of ensuring calculation accuracy; (ii) the divided zones provide an accurate judgment basis for the failure mode and damage state directly; (iii) for the reinforced concrete shield building (RCSB), the damage gradually expands upon a sudden change in structural stiffness at the upper junction of the reinforced concrete auxiliary building (RCAB); (iv) seismic fragility analysis of the RCAB presents the form of gradual reduction of damage from the upper layer to the lower layer.

Hydrogen production in solid oxide electrolyzers coupled with nuclear reactors

In this study, two types of high temperature electrolyzers (O=SOE and H+SOE) were investigated for hydrogen generation in relation to nuclear power plant operations. The analysis encompasses the thermal integration of proton and ion conducting solid oxide electrolyzers, which are fed with steam generated in the nuclear plant. Under consideration in the study was the steam turbine cycle of an AP1000 nuclear power plant. The main parameters of electrolysis were tailored to match the typical operating temperature of the electrolyzers, and the water utilization factor was set at the same value for the two technologies under consideration. There are some advantages to applying high temperature electrolysis to the deaerator steam feed: first, there is almost no modification of the nuclear steam turbine cycle; second, flexibility of the nuclear power plant rises by 20% with almost constant thermal load of the nuclear reactor; and third, high pressure hydrogen is obtained for commercial purposes. The analysis concludes that hydrogen can be produced in electrolyzers integrated with nuclear plants at an energy cost of 38.83 and 37.55 kWh kgH2−1 for protonic and ionic solid oxide electrolyzers, respectively.

Research on the impact effect of AP1000 shield building subjected to large commercial aircraft

This study addresses the numerical simulation of the shield building of an AP1000 nuclear power plant (NPP) subjected to a large commercial aircraft impact. First, a simplified finite element model (F.E. model) of the large commercial Boeing 737 MAX 8 aircraft is established. The F.E. model of the AP1000 shield building is constructed, which is a reasonably simplified reinforced concrete structure. The effectiveness of both F.E. models is verified by the classical Riera method and the impact test of a 1/7.5 scaled GE-J79 engine model. Then, based on the verified F.E. models, the entire impact process of the aircraft on the shield building is simulated by the missile-target interaction method (coupled method) and by the ANSYS/LS-DYNA software, which is at different initial impact velocities and impact heights. Finally, the laws and characteristics of the aircraft impact force, residual velocity, kinetic energy, concrete damage, axial reinforcement stress, and perforated size are analyzed in detail. The results show that all of them increase with the addition to the initial impact velocity. The first four are not very sensitive to the impact height. The engine impact mainly contributes to the peak impact force, and the peak impact force is six times higher than that in the first stage. With increasing initial impact velocity, the maximum aircraft impact force rises linearly. The range of the tension and pressure of the reinforcement axial stress changes with the impact height. The perforated size increases with increasing impact height. The radial perforation area is almost insensitive to the initial impact velocity and impact height. The research of this study can provide help for engineers in designing AP1000 shield buildings.

Special equipment for processing inner holes in bending section of the main pipeline in nuclear power plant

Because of the special forging material and complex pipe structure of the main pipe of the AP1000 nuclear power plant, a single general equipment can not meet the requirements of machining the inner hole of the bending section. At present, most of them adopt the method of approaching theoretical dimension by manual grinding, which makes the labor intensity of workers large, the work cycle is long, and it is difficult to ensure the machining accuracy of the workpiece. In view of the practical difficulties in processing the elbow inner hole of nuclear power main pipeline, the copying processing equipment is designed. The fixed circular arc guide rail is used in the equipment. The arc curvature of the guide rail is the same as that of the workpiece. The boring cutter head moves axially along the fixed circular arc curvature guide rail for copying boring. After testing, it fully meets the requirements of the shape, position and size accuracy of the inner hole in the bending section of the nuclear power main pipeline. The method is efficient, accurate and innovative, and provides reference for similar product processing schemes.

A methodology of risk assessment, management, and coping actions for nuclear power plant (NPP) hit by high-explosive warheads

A detailed Loss of Coolant Accident (LOCA) analysis in an AP1000 NPP was performed, followed by a definition of the vulnerability analysis principles, and analysis of blast loads and fragments impact created by a nearby explosion. The AP1000 NPP performs excellently to small-break LOCA due to in-structure shock, with the 10 CFR 50.46 Acceptance Criteria fully accomplished. Impulsive dynamic loads resulting from blast waves and fragments impact of GBU-28 (Guided Bomb Unit) were considered for a nearby explosion. We model the structure and the main reactor components using the MSC/Dytran code to obtain accurate internal acceleration levels at critical points. We account for the appropriate blast wave interaction with the soil and the soil interaction with the containment structure, rather than using empirical formulas. The model includes the shielding structure with its concrete base, the support structures for the reactor, the steam generators, and the pressurizer. The combined effect of bomb fragmentation and blast loading was also considered using a cylindrical fragmentation model and the blast model of Kingery-Bulmash, assuming a hemispherical charge. A comprehensive risk assessment methodology composed of four phases was developed. The methodology is comprised of: (I) System analysis, (II) Hazard analysis, (III) Damage assessment, and (IV) Risk analysis of the in-structure shock consequences. Using seismic fragility curves for analysis of the expected failure modes according to explosion events faced difficulties since no published data was found. Adjustments to these fragility curves were made using median acceleration limits on components designed to withstand airplane crash, together with standard deviations taken from the given earthquake fragility tables. The findings reveal that the probabilities of failure of the reactor coolant system components resulting from a GBU-28 nearby hit, namely the pressurizer, the cooling pumps, and valves are quite high (greater than 1∙10−4).

The development of high fidelity Steam Generator three dimensional thermal hydraulic coupling code: STAF-CT

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermal-hydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including “Transient” and “Fully Three Dimensional Coupling” features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

Global sensitivity analysis of the main steam line break accident by using sampling methods and surrogate models

Global sensitivity analysis (GSA) aims at quantifying the individual effects of input variables on the model outputs. The global sensitivity analysis is expressed by the Sobol’ indices. The Sobol’ indices can be obtained by different algorithms. These algorithms can be divided into two main categories: the sampling methods and the surrogate models. This work employs these methods to perform the global sensitivity analysis of Main Steam Line Break Accident (MSLB) of the AP1000 nuclear power plant. Specifically, the DAKOTA is coupled with an in-house containment analysis code to perform the sensitivity analysis. This study focuses on the effects of five key input parameters on safety parameters. The results show that atmospheric pressure has the most important impact on the peak pressure and peak temperature of the containment during the MSLB accident. Besides, compared with the sampling methods, the surrogate model methods need fewer samples and less time to obtain the Sobol’ indices.