Nuclear Power Plant Model Research Articles

Design and Development of Fractional Order Convolutional Neural Network Based Fractional Order Nonlinear Reactor Power Simulator for CANDU-PHWR

A highly complex nonlinear Reactor Regulating System (RRS) of Canadian Deuterium Uranium Pressurized Heavy Water Reactor (CANDU-PHWR) based Nuclear Power Plant (NPP) simulated in the present research. The internal design of RRS is secured and vendor controlled which is embedded in AC-132 Programmable Logic Controller (PLC). Therefore, the problem of the identification of the RRS controller model is addressed. A data-driven Fractional Order Nonlinear MIMO Hammerstein Model (FO-NC-MIMO-HM) of NPP is identified using an Adaptive Immune Algorithm (AIA) based on a Global Search Strategy (GSS) and Auxiliary Model Recursive Least Square Method (AMRLSM). Parameters of FO-MIMO-HM are identified using Innovative Real-Time Plant Operational Data (IRTPOD). The original PLC-based controller is replaced with a new Fractional Order Convolutional Neural Network (FO-CNN) based Fractional Order Nonlinear Controller (FO-NC). Therefore, a visual Simulator is developed for detailed modeling, control, simulation, and analysis of the proposed design scheme for RRS in Visual Basic (VB) Software. The performance of the proposed design scheme is tested and validated for different modes of RRS against benchmark data obtained from Plant Data Recorder (PDR) and found in close agreement well within the design bounds.

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.

Risk-informed Emergency Response Training for Backdraft in Nuclear Power Plants

Research has been conducted for developing fire evacuation and response training programs for nuclear power plant (NPP) application. Among numerous fire scenarios that may occur in an NPP environment, three different points of origin for a fire were selected for the program based on a risk-informed approach: switchgear room, main control room, and safety injection pump room. Fire outcomes were predicted for these scenarios via numerical modeling and the results were incorporated into the newly developed fire evacuation and response training program for the APR1400, Korea’s next-generation NPP model. The switchgear room fire scenario was found to have the most potential for backdraft to occur during manual fire response following automatic gaseous fire suppression system activation. The emergency response manual does discuss this possible backdraft occurrence; however, the guidance to avoid injuries is qualitative, such as to be cautious of backdrafts and wait a sufficient amount of time after opening a door before entering the. In this study, backdraft phenomenon that may occur from a switchgear room fire was numerically examined using the recent version of the Fire Dynamics Simulator to develop an appropriate timeline to be implemented in the fire evacuation and response training program. Based on the findings, the following guidance is provided: (1) backdraft can only occur when the fire originates in the space near the door; (2) wait at least 10 minutes after opening the door before entering the room; (3) watch for rapid smoke production, as this may be an antecedent phenomenon of backdraft; and (4) when smoke production increases rapidly, leave the room as soon as possible to avoid being caught within the deflagrating flames from a backdraft.

Global torsional response of seismically isolated nuclear plants and evaluation of codal provisions

A major challenge in the application of seismic isolation to nuclear structures is the torsional response of base-isolated Nuclear Power plants (NPPs), which amplifies the isolator displacement around the periphery of the isolation system. Additionally, degradation in the mechanical properties of the isolator due to the response associated with extreme ground shaking may also affect the torsional response. This paper presents a comprehensive study on the torsional response of a NPP seismically isolated using lead rubber bearings with a focus on the isolation properties and eccentricity in the isolation system. Nonlinear response history analysis is conducted with different NPP model representations, which are derived from a benchmark NPP model for a range of eccentricities with different time periods and characteristic strengths. It was observed that torsional response in base-isolated NPPs could be controlled by using a flexible isolation system of higher characteristic strength, and it may be important to include the strength degradation in lead rubber isolators to correctly estimate torsional amplification to isolator displacements. The torsional amplification predicted by the nonlinear response history analysis was also compared with the provisions of ASCE 7 for buildings, and recommendations are provided on their suitability for seismically isolated NPPs.

Analysis of the Influence of Structure–Soil–Structure Interaction on the Seismic Response for Nuclear Power Plant ASTS

After the Fukushima nuclear accident in Japan, whether the structural safety of the nuclear power plant (NPP) can be guaranteed under an earthquake has been of wide concern. Automatic seismic trip systems (ASTS) have been deployed in NPPs. There are generally two units or more double units on the one NPP site, and the vibration energy of the structure can certainly affect its adjacent structures through the soil, and there is energy transfer and conversion between adjacent structures. At present, the distance between two reactors of NPPs of different reactor types is generally 100–200 m (this distance is referring to the distance between the centers of two reactors, which is slightly different for different reactor types). In the past, seismic instruments were used in the ASTS and the setting of shutdown threshold, so the impact of only one unit was considered in the structural analysis of NPPs, and the interaction of two reactors through site conditions was not considered. In addition, as the site conditions of NPPs become more and more complex, it is necessary to consider the impact of one reactor structure on another reactor structure through the soil on the same site under an earthquake. In order to analyze the influence of structure–soil–structure interaction (SSSI), a three-dimensional refined finite element model of NPPs is established in this paper. The soil–structure interaction (SSI) is considered by using viscoelastic boundary. The seismic responses of different positions of the raft foundation and NPP structure, under the conditions of considering SSI effect (one reactor) and SSSI effect (two reactors), are compared. The influence of SSSI effect on the seismic responses of the raft foundation and NPP structure is revealed. It is proposed that SSSI effect should be reasonably considered according to the site conditions for the structural analysis of NPPs for the NPP ASTS.

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.

L₁-Adaptive Robust Control Design for a Pressurized Water-Type Nuclear Power Plant

This work proposes adaptive control-based design strategies to control a pressurized water reactor (PWR) nuclear power plant (NPP). An L <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">1</sub> -adaptive-based state-feedback control technique is proposed using the linear quadratic Gaussian control and projection-based adaptation laws. The control scheme possesses good robustness capabilities in handling disturbances and uncertainties. A robust L <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">1</sub> -adaptive control technique is also proposed by combining the L <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">1</sub> -adaptive control with the loop transfer recovery (LTR) technology. The framework hence gives the strengthened robust set-point tracking performance given the matched and unmatched uncertainties and disturbances. The NPP model employed in this article is defined by five inputs, five outputs, and 38 state variables. A linear model for controller design is obtained by linearizing the nonlinear NPP model at operating conditions. Various simulations are carried out on subsystems of the NPP to verify the effectiveness of the proposed scheme. Numerical and statistical measures are computed for quantitative analysis of the controllers' performance. Several classical control design techniques are also implemented, and their performance is compared with the proposed adaptive control techniques.

Damage and vibrations of nuclear power plant buildings subjected to aircraft crash part II: Numerical simulations

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to numerically assess the damage and vibrations of NPP buildings subjected to aircrafts crash. In Part I of present paper, two shots of reduce-scaled model test of aircraft impact on NPP were conducted based on the large rocket sled loading test platform. In the present part, the numerical simulations of both scaled and prototype aircraft impact on NPP buildings are further performed by adopting the commercial program LS-DYNA. Firstly, the refined finite element (FE) models of both scaled aircraft and NPP models in Part I are established, and the model impact test is numerically simulated. The validities of the adopted numerical algorithm, constitutive model and the corresponding parameters are verified based on the experimental NPP model damages and accelerations. Then, the refined simulations of prototype A380 aircraft impact on a hypothetical NPP building are further carried out. It indicates that the NPP building can totally withstand the impact of A380 at a velocity of 150 m/s, while the accompanied intensive vibrations may still lead to different levels of damage on the nuclear related equipment. Referring to the guideline NEI07-13, a maximum acceleration contour is plotted and the shock damage propagation distances under aircraft impact are assessed, which indicates that the nuclear equipment located within 11.5 m from the impact point may endure malfunction. Finally, by respectively considering the rigid and deformable impacts mainly induced by aircraft engine and fuselage, an improved Riera function is proposed to predict the impact force of aircraft A380.

Damage and vibrations of nuclear power plant buildings subjected to aircraft crash part I: Model test

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to experimentally assess the damage and vibrations of NPP buildings subjected to aircraft crash. In present Part I, two shots of reduce-scaled model test of aircraft impacting on NPP building were carried out. Firstly, the 1:15 aircraft model (weighs 135 kg) and RC NPP model (weighs about 70 t) are designed and prepared. Then, based on the large rocket sled loading test platform, the aircraft models were accelerated to impact perpendicularly on the two sides of NPP model, i.e., containment and auxiliary buildings, with a velocity of about 170 m/s. The strain-time histories of rebars within the impact area and acceleration-time histories of each floor of NPP model are derived from the pre-arranged twenty-one strain gauges and twenty tri-axial accelerometers, and the whole impact processes were recorded by three high-speed cameras. The local penetration and perforation failure modes occurred respectively in the collision scenarios of containment and auxiliary buildings, and some suggestions for the NPP design are given. The maximum acceleration in the 1:15 scaled tests is 1785.73 g, and thus the corresponding maximum resultant acceleration in a prototype impact might be about 119 g, which poses a potential threat to the nuclear equipment. Furthermore, it was found that the nonlinear decrease of vibrations along the height was well reflected by the variations of both the maximum resultant vibrations and Cumulative Absolute Velocity (CAV). The present experimental work on the damage and dynamic responses of NPP structure under aircraft impact is firstly presented, which could provide a benchmark basis for further safety assessments of prototype NPP structure as well as inner systems and components against aircraft crash.

Damping implementation issues for in‐structure response estimation of seismically isolated nuclear structures

AbstractSeismic (base) isolation of a nuclear power plant (NPP) is a viable strategy to reduce seismic demands in structural elements and nonstructural systems and components. Accurate estimation of in‐structure response of base‐isolated NPPs is critical to the analysis and design of safety‐related structural components. Damping models and their implementation in contemporary analysis platforms play a significant role in the numerical response estimation of a nonlinear system. Most of the recent damping studies cater to moment frame buildings for which nonlinearities are modeled using massless rotational springs. There are unique challenges associated with the implementation of damping to numerical models of the base‐isolated NPPs due to the broad frequency range of interest and the markedly different damping characteristics of the isolators and the superstructure. In‐structure response spectra of a base‐isolated NPP are sensitive to the chosen numerical and modeling techniques, both of which are investigated in this paper. Three representations of base‐isolated NPP are considered: (1) a two‐node macro model, (2) a lumped‐mass stick model, and (3) a detailed finite element model. Rayleigh damping, modal damping, uniform frequency damping, and several combinations thereof were used with the three NPP models to obtain the median response of the base‐isolated NPP subject to 30 sets of three‐dimensional ground excitations. The results allow informed decisions regarding the selection of a numerical model and damping implementation to different analysis platforms needed to obtain a reliable in‐structure response of base‐isolated NPPs subject to earthquake shaking.

Damage development analysis of the whole nuclear power plant of AP1000 type under strong Main-aftershock sequences

This paper focuses on the seismic damage analysis of the whole nuclear power plant (NPP) of AP1000 type during strong seismic sequences. For that purpose, seven synthetic main-aftershocks with three-direction inputs are considered in this study, and the three-dimensional model of whole NPP of AP1000 type suffering from the selected seismic sequences is established. The results indicate that (i) the damage aggravation effect is substantially caused by considering the reinforced concrete auxiliary building (RCAB) for the NPP and is mainly concentrated at the upper intersection between the reinforced concrete shield building (RCSB) and RCAB, especially during beyond design basis earthquake sequences, (ii) concrete damage are greatly developed with increasing PGA, and aggravated by the aftershock, concrete tension damage is more severe than compression damage under earthquake sequences, (iii) the structure residual displacement in three directions are increased by the aftershock, (iv) the average value of the sequence seismic damage dissipation energy is 1.14–1.27 times that of a single earthquake, and the increasing trend of the times is linear.

Detailed and verified NPP model for safe interconnection with electrical grid

A precise understanding for nuclear power plant (NPP) response to grid disturbances is required to ensure safe and reliable operation for both NPP and power grid. NPP performance is studied by dynamic modeling and simulation, since it is not practical to study it by experiments. In this work, a mathematical model for complete pressurized water reactor NPP is represented based on MATLAB/Simulink. First, the main parts of the model (reactor, steam generator, pressurizer, and steam turbine) are studied and verified individually. Then, those parts are connected together in a highly controlled system to test the overall plant behavior under load changes. The results of simulation indicate that the plant is capable to regulate its output according to electrical grid demand and reach the steady-state conditions. Thus, a high degree of safe interface between the electric grid and NPP is achieved.

The future role of nuclear power in the coal dominated power system: The case of Shandong

Nuclear energy and renewable energy have been deployed and expanded globally to reduce greenhouse gas emissions and improve the environment. The demand for flexibility is increasing sharply with the rapid penetration of intermittent renewables in the power system. Meanwhile, in the coal dominated power system, the reduction on the share of coal-fired power will leave the power system with a more serious shortage of flexibility. The flexible nuclear operation has aroused wide attention as an important flexibility resource of the power system. To evaluate the role of the flexible nuclear operation on the coal-dominated power system, we first model the flexible operation of nuclear power plants (NPPs). Then we integrate the operation model of NPPs in a unit commitment and economic dispatch model. The model is further integrated into the power system operation simulation framework. Finally, we analyze the potential benefits of flexible nuclear operation using the power system operation simulation tool for a coal-dominated power system: Shandong province of China. The simulation results indicate that the flexible nuclear operation could effectively reduce start-up and shut-down times of coal-fired power units and help the power system to achieve a higher renewable energy penetration of 47.98%. The research suggested that it is feasible and beneficial to use NPPs to produce flexibility in the future power system.

Advanced method for neutronics and system code coupling RELAP, PARCS, and MATLAB for instrumentation and control assessment

The operation of current nuclear power plants (NPP) and the deployment of advanced nuclear reactors rely on digital instrumentation and control (I&C) and information technology systems. The growing use of digital systems implies considerable security and safety challenges, as the usual standard codes used for designing, licensing, and analyzing accidents over the last few decades have been limited in terms of I&C features. This study presents an innovative method to couple PARCS/RELAP5 with the high-performance language MATLAB/Simulink. As a proof-of-concept proof, a 2772 MWt pressurized water reactor modeled using the coupled package PARCS/RELAP5 is simulated. A supervisory system is implemented that allows MATLAB/Simulink to oversee the nuclear codes for exchanging online data. In addition, I&C high-level computing tools are added to the NPP model, endorsing the application of “on-the-fly” decision logics. The coding strategy does not change the nuclear binary source codes, as the PARCS/RELAP5 codes have been validated for design and licensing. By running open- and closed-loop cases for control rod assembly movement, we found that PARCS/RELAP5 capabilities for digital I&C modeling and simulations can be extended by adding MATLAB/Simulink flexibility.

Perfecting the use of hybrid models in scaling analysis

Different methodologies devoted to qualify Nuclear Power Plant (NPP) system-code nodalizations rely on Kv-scaled calculations as an essential tool for their purpose. In the framework of Power-To-Volume strategy, a Kv-scaled calculation is a simulation in which, defined Integral Test Facility (ITF) test conditions are scaled-up to a NPP nodalization to reproduce the same scenario. In the recent years, the “Universitat Politècnica de Catalunya” (UPC) has developed SCUP, a scaling methodology to qualify NPP models. Following the steps established in UPC’s SCUP methodology, Kv-scaled calculations become really useful when two different kinds of them, the so-called “pure scaled” and “hybrid” are properly combined.This paper starts summarizing the most relevant features of UPC’s SCUP methodology and its validation process. It also includes important aspects of the usefulness of hybrid calculations in their role of explaining distortions that may appear when a plant simulation is compared with a scaled-up version of test results.Some ideas on the forthcoming uses of hybrid nodalizations are also presented. Among such uses the most significant is the support to test design. Hybrid calculations are used to justify facility modifications. This particular, despite being more specific for future modular ITFs, is illustrated in the paper for minor changes in existing facilities.

The core shroud leakage analysis and study for Kuosheng nuclear power plant

Abstract Kuosheng Nuclear Power Plant (NPP) is a BWR/6 plant in Taiwan. First, this study focuses on the establishment of the TRACE/SNAP model for Kuosheng. To check the system response of the Kuosheng TRACE/SNAP model, the Final Safety Analysis Report (FSAR) and startup tests data are used to assess the TRACE/SNAP model. The TRACE predictions are consistent with the FSAR and startup tests data. This indicates that there is a respectable accuracy in the TRACE/SNAP model of Kuosheng NPP. Second, this study also focuses on the application of the TRACE/SNAP model in the core shroud leakage. The Kuosheng NPP TRACE/SNAP model is used to perform the analysis of core shroud leakage. The TRACE results imply that Kuosheng NPP is in a safe situation when the core shroud leakage transient occurs.

Simulation of SB-LOCA of typical PWR with MELCOR code

Safety of Nuclear Power Plants (NPPs) is one of the main issues in nuclear industry mainly because of the probability of radionuclides release to the environment. Due to this fact nuclear safety is continuously being improved by vendors to face the growing demands especially after the accident in Fukushima. This accident because of its character (station blackout scenario) resulted in putting more emphasis than ever on passive safety systems in NPPs. Such systems are one of main focus points in presented simulation. One of the ways to improve NPPs’ safety and check its robustness is to proceed simulations of different events and find the weakest points inside the system. First thing is to create simulation model of NPP which has to be validated in order to prove reliability of future results. This paper presents simulation run of design basis accident for large PWR reactor based on AP1000. Model was developed using publicly available data only, which means that it should be considered only as “AP1000-like model”. MELCOR code which was used in presented calculations is a severe accident code which simulates core melting during the worst possible accidents in NPP. Here it was decided to simulate design basis accident namely small break loss of coolant accident (SB-LOCA) and compare obtained results with those from the safety reports. Main motivation to create such simulation was to increase authors’ knowledge about MELCOR behaviour and sensitivity in case of thermal hydraulic response to learn what kind of errors and uncertainty could be expect. Second motivation was to check if Melcor is capable to simulate design basis accident with acceptable results. Finally the obtained results are in relatively good agreement with the expectations and some differences that have appeared do not affect general course of the accident. Shape of transient curves and timing of events are reasonable correct.

Evaluation of control room habitability in case of LOCA for Maanshan NPP using codes RADTRAD, HABIT and ALOHA

Abstract The method for the evaluation of the control room habitability is presented in this paper with focus on Maanshan PWR nuclear power plant (NPP) using the codes RADTRAD, HABIT, and ALOHA. Therefore, this paper is divided into two parts: The first part is the evaluation of the cumulative dose at the control room, the exclusion area boundary (EAB) and the low population zone (LPZ) in case of an design basis loss of coolant accident (DBA/LOCA). For this first part, the Maanshan NPP models of the code RADTRAD/SNAP were used for the analysis. The second part is the evaluation of the control room habitability under the assumption of CO2 storage burst. For this part, the HABIT and ALOHA codes were used. As result it was seen that the RADTRAD calculation results are below the failure criteria of standard review plan (SRP) and 10 CFR 100.11. The HABIT and ALOHA results are below the R.G. 1.78 failure criteria. These results indicate that Maanshan NPP’ habitability can be maintained under the above conditions.

Seismic response of base isolated nuclear power plants considering impact to moat walls

Seismic isolation can be an effective strategy to protect critical facilities including Nuclear Power Plants (NPPs) from the damaging effects of horizontal earthquake ground shaking. The increased flexibility at the base and resulting elongation of the natural vibration period of the structure leads to significant reductions in forces and acceleration transmitted to the structure above the isolation level at the expense of displacements concentrated in the isolation system. Displacement demands can be large at sites of moderate to high seismic hazard and can be accommodated by a horizontal clearance or moat at the isolation level, typically located in the basement of a structure. A surrounding moat wall can function as a stop to limit isolation system displacements and prevent bearing failure under beyond design basis shaking. However, impact of the isolated structure against the moat wall is of concern due to potential amplification in response of the superstructure. Design guidelines aim to prevent impact by specifying a required minimum clearance to stop (CS) with a low annual frequency of exceedance. A CS below the required value can be justified through analysis considering impact to the moat wall or stop. However, little guidance is available on how to model impact to the moat wall and resulting effects on the NPP superstructure. This study proposes a simplified model for impact simulation that captures the impact forces and the effects of impact on the response of seismically isolated NPPs. Variable clearance to the stop and a range of properties for the moat wall and isolation system are considered to identify parameters that influence the response. Results of these studies indicate that large NPP plants as considered here can have significant penetration into the moat wall, and thus not fully limit displacements in the isolation system, while having considerable increases in accelerations throughout the height of the NPP model.