Coolant Accident Research Articles

A New Experimental Method for the Nonlinear Modal Parameter Identification of a Pressurized Water Reactor Fuel Assembly

Establishing a dynamic model that accurately describes a realistic pressurized water reactor (PWR) fuel assembly is crucial to precisely evaluate the mechanical properties of the fuel assembly in seismic or loss of coolant accidents (LOCAs). The pluck test combined with the logarithmic decrement method has been widely applied in previous studies to extract fundamental modal parameters to calibrate dynamic models. However, most previous investigations focused on the first cycle of free vibration, which is strongly affected by stiction, baseline shift, drop conditions, and high-order mode interference, leading to inaccurate results. Moreover, these traditional methods cannot be used to extract high-order modal parameters. In this work, a novel experimental method for identifying the nonlinear modal parameters of a PWR fuel assembly is proposed. First, two algorithms are adopted to decompose the free vibration. Second, the local linearized modal parameters are extracted by a single-degree-of-freedom fitting method with a sliding window. Finally, these local linearized modal parameters are summed to obtain the nonlinear relationships between the modal parameters and amplitude. The new method makes more effective use of experimental data, obtains more accurate modal parameters than the logarithmic decrement method, and is capable of extracting high-order modal parameters. In the end, the test results are fitted by a fractional polynomial, which is of great value for numerical simulations.

Investigation of the source term and radiological consequences of small break LOCA with failure of the emergency core cooling system in Bushehr nuclear power plant

In the present study, the in-containment source term derivation, environmental dispersion and deposition as well as radiological consequences of Bushehr Nuclear Power Plant (BNPP) due to Small Break Loss of Coolant Accident (SB-LOCA) have been studied. First, the calculation of in-containment release and distribution of the source terms due to the severe accident of SB-LOCA with failure of the active part of the Emergency Core Cooling System (ECCS) has been carried out. In this calculation, two scenarios, including with and without operator control actions, have been considered. Then, the environmental leakage of radionuclides as well as their dispersion and deposition considering two meteorological scenarios have been investigated. Consequently, public exposure has been calculated for each meteorological scenario and the obtained results have been reported and discussed. For two accidents with the same meteorological conditions, at a distance of 400 m from the release point, the received dose in the first 24 h is 18 and 342 mSv without and with operator action, respectively. The results show that due to the intervention of the operator and the use of emergency cooling supply systems, more water enters the reactor. Therefore, if the operator's actions do not prevent the core from melting, more source terms have entered the containment after the core melts and worse conditions than the baseline scenario without operator action occurred.

Thermal-hydraulic optimization of a proposed EU-DEMO hydrogen passive removal system

As the R&D of magnetic fusion power demonstrating plants are approaching important steps toward concept designs, analysts are working parallelly on the safety assessment of such concepts to identify any potential risk. One of the safety concerns involves the confinement of radioactive substances during normal operation and accidental conditions. Several accident sequences inside the tokamak vacuum vessel or pressure suppression systems are characterized by the risk of hydrogen buildup and subsequent ignition that could threaten the structural confinement integrity. In the Safety and Environment work package of the EUROfusion consortium, possible approaches to mitigate the hydrogen explosion risk are under investigation. One of the exploratory solutions is based on limiting the hydrogen concentration that could reach flammable gas mixture conditions and using Passive Autocatalytic Recombiners (PARs) installed into the atmosphere of the pressure suppression systems tanks to recombine hydrogen.This paper examines the theoretical effectiveness of the PARs intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled Lithium Lead (WCLL) concept of EU-DEMO, using an optimization methodology. The involved systems have been modelled in MELCOR to estimate the PARs recombination capability as a function of the thermal-hydraulic parameters of the suppression tanks. Furthermore, the optimizer entity of the RAVEN tool is applied to perform optimization studies on the hydrogen recombination system design parameters. The goal is to explore the geometrical and thermal-hydraulic parameters that maximize the capability of the hydrogen removal system for the WCLL concept.

Spray Cooling of Vertical Inline Horizontal Tubes

In water-cooled nuclear reactor safety, quench cooling is essential for minimizing core damage in the event of a Loss of Coolant Accident (LOCA). A hot, dry surface is quenched by bringing it into abrupt contact with a coolant, hence causing a fast fall in surface temperature. In this study, an air-atomized water spray cooling facility was built, and a series of tests were carried out to explore the quenching and rewetting of vertical inline horizontal tubes. During the cooling process, constant heat flux was provided on each tube. In the absence of a coolant, the steady heat flux represents the decay power of the channel components that reach the calandria tube surface. Temperature variations in the circumferential direction were measured over time during cooling. The effect of water pressure and air pressure on the rapid cooling of vertical inline horizontal tubes has been investigated. As air pressure rises, the quenching temperature of each tube decreases. A high-speed video camera was employed to investigate the flow behavior on the top tube surface as well as the spread of the re-wetting front around it. Experiments and visual analysis show that wetting front velocity rises with the water flow rate.

Fluidelastic instability of a rod bundle subjected to jet cross-flow

Jet flow emanating from loss of coolant accident (LOCA) holes in the baffle plates of some pressurized water reactors (PWRs) was found responsible for flow-induced vibration in nuclear fuel bundles. This could cause fretting wear of the fuel rods at the spacer grid location. Experimental tests on prototypical rod bundles subjected to jet cross-flow showed a strong vibration response akin to fluidelastic instability of tube arrays in uniform cross-flow. Vibrations similar to flow-periodicity resonance response have also been reported.Fluidelastic instability (FEI) is the most critical flow-induced vibration mechanism. For fuel rods subject to jet-cross flow, it is important to identify the specific mechanism underlying the rod vibrations and in particular, confirm possible fluidelastic instability. In previous work, a few stability boundary correlations based on the Connors equation were developed for the baffle jetting (i.e. 2D jet) problem in PWRs. With the exception of these correlations, no theoretical model accounting for fluid–structure interaction under jet cross-flow has been reported.To investigate possible fluidelastic instability, the quasi-steady model, initially developed for tube arrays in uniform cross-flow, is extended to the problem of a rod bundle subjected to a jet cross-flow. In the extended quasi-steady model, a new formulation of the fluidelastic forces as functions of the projected rod area derivative is proposed to account for the variation of the projected area through the LOCA hole (i.e. circular jet).A significant limitation of the ‘classical’ quasi-steady model is the poor representation of the time delay effect, known to be critical for correct fluidelastic instability prediction. This shortcoming is addressed in the present work. Furthermore, unsteady flow effects are partly taken into account experimentally measuring the time delay parameter for this jet-cross-flow configuration.For the present highly compact rod array, a new test apparatus is designed and built to directly measure the time delay parameter. The new results show that the time delay to be non-constant, strongly dependent on Reynolds number while relative independent of reduced velocity for Low Re. For High Re, the time delay depends on both Re and reduced flow velocity. Based on the new measurements, a formulation for the time delay parameter, dependent on Reynolds number and reduced velocity is proposed for the highly packed fuel rod bundle.The new extended quasi-steady model conclusively confirms that the vibration phenomenon observed in experimental tests is clear fluidelastic instability induced by the jet cross-flow. Moreover, the model predicts the critical jet velocity for fluidelastic instability with an error of 9% when the new Reynolds number dependent time delay formulation is employed.

Exploratory study of the EU-DEMO Water-Cooled Lithium Lead breeding blanket behaviour in case of loss of cooling capability

Within the framework of the European Roadmap to the realization of fusion energy, a strong international cooperation is ongoing to develop a Breeding Blanket (BB) system for the EU-DEMO reactor. Although it is still to be decided whether the DEMO in-vessel components should perform any safety function, the pursuing of robust blanket concepts able to handle upset and accidental loading conditions has been always seen as good practice in fusion reactor engineering to enhance the inherent plant safety performances. Amongst the several classes of events that might challenge the BB structural integrity, the large Loss of Coolant Accident is one of the most relevant because it usually leads to a fast loss of cooling capability of the structures. Due to the characteristic of the tokamak assembly, the behaviour of each blanket segment during a sudden loss of cooling capability does not depend only upon distinguishing features of the component itself. In fact, the overall transient can be governed by conditions established in surrounding elements, like adjacent blanket segments and vacuum vessel, as well as by the plasma shutdown strategies adopted to protect the reactor.The scope of the activity herein presented is to make a preliminary assessment of the intrinsic capability of EU-DEMO tokamak architecture to cope with the loss of cooling in the Water-Cooled Lithium Lead (WCLL) BB concept. Evaluation of BB thermal field in short and medium term under simplified, yet conservative, assumptions was carried out for four transient scenarios with the aim of investigating the response of the structure in case of: a) fast or soft plasma shutdown, and b) different blanket cooling schemes. Moreover, the WCLL BB thermo-mechanical response in the most critical time steps has been assessed. The obtained results shall help for future decisions on safety systems/action to be implemented to cope with accidents.

Post-LOCA control room dose analysis for Maanshan NPP using the AST methodology

Abstract To address the issue identified in USNRC’s Generic Letter (GL) 2003-01 that the unfiltered air in-leakage rate through plant’s control room envelope during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room envelope unfiltered in-leakage Tracer Gas Test (TGT) for Maanshan nuclear power plant (NPP) has been performed in 2020. For future applications, an improved control room dose analysis approach using Alternative Source Terms (AST) has been developed in present study to check whether the TGT results meet regulatory limits and have sufficient safety margins. The AST method follows Regulatory Guide 1.183 (RG 1.183) and the TGT results must be fulfilled the total effective dose criteria set forth in 10 CFR 50.67. Based on the AST approach and the TGT results, the control room personnel dose for Maanshan NPP during Loss of Coolant Accident (LOCA) is 14.35 mSv, and the safety margin is 71.3%. It is sufficient to cover the effects of structural ageing and changes in meteorological data during the control room habitability reassessment and the analysis uncertainty.

Analysis of nuclear accident scenarios and emergency planning zones for a proposed Advanced Power Reactor 1400 (APR1400)

Radiological scenario analysis considering potential accidents and the corresponding emergency planning arrangements are based on the estimations of radioactive releases following such accidents. This involves investigating the possible effects on the surrounding population and environment. The current paper hypothesized three accident scenarios in a pressurized water reactor (PWR) proposed to be located in Khor Duweihin on the Western Coast of the Arabian Gulf. HotSpot tool was used to estimate the total effective dose equivalent (TEDE) due to certain released radionuclides of different chemical groups. Initial emergency planning arrangements were defined based on the estimated doses triggered by the hypothetical scenarios considered: loss of coolant accident (LOCA), station blackout (SBO) and steam generator tube rupture (SGTR) in an Advanced Power Reactor 1400 (APR1400). This study established a baseline for an emergency plan that assures the mitigation of radiological consequences due to probable nuclear accidents, involving preparedness based on forecasting the atmospheric dispersion of radionuclides released during potential scenarios in different meteorological conditions.

Using TRACE to establish the analysis model of Kuosheng nuclear power plant for decommissioning transition phase

Abstract The Unit 1 of Kuosheng nuclear power plant is in the decommissioning transition phase and still has spent fuel in the core of reactor now. The time to reach the TAF (top of active fuel) for water level and reach 600/800 °C for cladding temperature are the key parameters in the safety analysis, these affect that the plant has how much time to handle transients. Therefore, to study the water level and cladding temperature for station blackout (SBO) and loss of coolant accident (LOCA) transients in the decommissioning transition phase, the analysis model of Kuosheng nuclear power plant was established by using TRACE code. To evaluate the effect of the decay heat of spent fuel on the water level and cladding temperature, the sensitivity analysis of decay heat was also performed in this study.

A Future with Machine Learning: Review of Condition Assessment of Structures and Mechanical Systems in Nuclear Facilities

The nuclear industry is exploring applications of Artificial Intelligence (AI), including autonomous control and management of reactors and components. A condition assessment framework that utilizes AI and sensor data is an important part of such an autonomous control system. A nuclear power plant has various structures, systems, and components (SSCs) such as piping-equipment that carries coolant to the reactor. Piping systems can degrade over time because of flow-accelerated corrosion and erosion. Any cracks and leakages can cause loss of coolant accident (LOCA). The current industry standards for conducting maintenance of vital SSCs can be time and cost-intensive. AI can play a greater role in the condition assessment and can be extended to recognize concrete degradation (chloride-induced damage and alkali–silica reaction) before cracks develop. This paper reviews developments in condition assessment and AI applications of structural and mechanical systems. The applicability of existing techniques to nuclear systems is somewhat limited because its response requires characterization of high and low-frequency vibration modes, whereas previous studies focus on systems where a single vibration mode can define the degraded state. Data assimilation and storage is another challenging aspect of autonomous control. Advances in AI and data mining world can help to address these challenges.

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO.

Investigation on damage assessment of fiber-reinforced prestressed concrete containment under temperature and subsequent internal pressure

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

Analysis of RELAP5 prediction of countercurrent flow limitation in downcomer at upper plenum test facility

The relevance of the countercurrent flow limitation phenomenon is known in nuclear (emergency core cooling and accident situations) industries, such as a Loss of Coolant Accident (LOCA). The occurrence of countercurrent flow limitation (CCFL), which can exist in a host of places around the nuclear reactor system, will prevent the emergency cooling water, which is necessary to cool the core successfully flowing into the core from the downcomer of Pressurized Water Reactor (PWR) in the LOCA. Accurate prediction of the downcomer CCFL phenomenon is significant for the design and analysis of reactor safety. Among the experiments at Upper Plenum Test Facility (UPTF), various tests, test 6 and test 7 are studied in this research. Those tests are conservatively evaluated to predict the CCFL phenomenon at the downcomer. Above all, the methodology of node partition used in the RELAP5 model for the channels with cross-connections to represent the circumferential downcomer annulus was also studied. In particular, the nodes along the circumferential direction had a vital effect on the prediction of the steam velocity profile of downcomers. The influence of the CCFL model, including the most widely used Wallis-type and Kutateladze-type correlations, and the cooling-limited liquid-downflow behavior were analyzed. In addition, the pressure on the downcomer and steam flow rate of the break were monitored, and the predicted results were also analyzed.

A deep learning-based framework for the operation prediction of primary heat transfer loop in nuclear power plants

A deep learning-based multi-node framework is constructed in this work to provide a data-driven platform that provides predictions for the operation condition of the primary heat transfer (PHT) loop in nuclear power plants (NPPs). Several deep learning models that have been verified and demonstrated in previous researches, such as Long-Short Term Memory (LSTM), Convolutional Neural Network (CNN), and zigmoid-based LSTM (zLSTM), are applied to modeling critical system parameters at three important nodes in the PHT loop. The feature extraction and process memory are enhanced via the collaborative work of CNN and LSTM. zLSTM, on the other hand, is successfully utilized to strengthen the long-term memory, especially for predictions of a node with multivariate inputs such as the steam generator. The node prediction results are also adopted for a polynomial fitting that generates an additional input to the next node, allowing each node to select a more accurate input. According to the verification experiments based on Loss of Coolant Accident (LOCA), the Mean Squared Error (MSE) result (1.29 × 10−3) and the Mean Absolute Error (MAE) result (1.37 × 10−2) of 0.7 cm2 LOCA case demonstrate the functionality and accuracy of the proposed framework. It is found that the fitting error (MSE) in the outlet node at 0.7 cm2 case is 38.5% lower than the prediction, showing the advantage of applying both deep learning and fitting methods. The best performance, in term of MSE, is obtained at SG node in the 0.7 cm2 case, where its processing error (0.001285) is 93.2% lower than that of the baseline models. Both the validation and verification experiments successfully proved the feasibility and advantages of the proposed framework, which offers an alternative option for the operation analysis of PHT performance.

Dust produced by plasma off normal event in the vacuum vessel: Experimental analyses of the deposition inside the pressure suppression tank during a LOCA

This paper deals with experimental tests simulating the dust deposition, produced in the Vacuum Vessel of the nuclear fusion reactor ITER, and carried by steam and gases inside the Pressure Suppression Tanks during a Loss of Coolant Accident (LOCA). The dust, mainly tungsten and beryllium, is produced by the erosion of the divertor or of the first wall due to plasma off-normal events, like plasma disruptions or vertical displacement events. The activity is performed in the frame of a research program performed at the University of Pisa, funded by the ITER Organization, for studying the behaviour of the Vacuum Vessel Pressure Suppression System (VVPSS). A reduced scale of VVPSS was built at the University of Pisa (1/32 geometrical scale). The experimental rig is composed by an electric steam generator and steam and air mass flow rate control lines. Steam with (or without) air is sent in a condensation container, partially filled with water. These fluids entrain a mass of dust measured by a dust mini-dosing system. The air and the non-condensed steam flow in the atmosphere through a demister and HEPA filters. A tests campaign was carried out injecting a simulant of the actual dust (aluminium oxide) in a flux of steam and air into the condensation tank. The tests were performed at different values of steam, air and dust mass flow rate. During the test, the dust concentration in the water was measured versus time, in order to determine the settling time. The dust deposition on the surfaces of the condensation tank was measured as well as the decontamination efficiency of the tank. The results demonstrated that about 99 % of dust is deposited on the tank surfaces and only a quantity less than 1 % is carried by the water during the tank discharge.

Principal component analysis of vibration response of a rod bundle subjected to jet cross-flow

Flow injecting through loss of coolant accident (LOCA) holes in the baffle plates of some pressurized water reactors (PWRs) can potentially induce vibrations of reactor fuel assemblies. For reactor fuel rods in close proximity to LOCA holes, the jet flow transverse to the rod axis may lead to vibration levels significantly higher than those for rods located away from the LOCA holes. The jet cross-flow-induced vibration could, in turn, induce fretting wear of the fuel rod cladding at the fuel assembly supports, also known as spacer grids.Fuel rods have axisymmetric flexibility, testing the real condition of the fuel assembly is important to know the most dominant vibration direction relative to the jet flow. This information could be used to suppress the dominant vibration by increasing the stiffness of spacer grids in the dominant direction. In this study, the effect of jet cross-flow on a 6x6 square rod bundle axisymmetric vibration is experimentally investigated. This research aims to answer the following important question: Does there exist a generalized eigenvector model that could be used to predict rod bundle vibration in jet flows, in addition to characterizing the stability effect of the diameter ratio? The dynamical behavior of the rod bundle is studied for three jet-to-rod diameter ratios; 1.6, 2.3, and 3.0. Flow-induced vibration tests are performed to determine the stability threshold of the bundle as well as to extract the bundle vibration mode shapes for the three nozzles tested. The experimentally measured mode shapes identify the dominant vibration direction which provides valuable insight into the fluid–structure interaction dynamics. The experiments show that increasing the diameter ratio lowers the critical velocity at which fluidelastic instability occurs. The maximum vibration amplitude in the bundle, on the other hand, decreases as the diameter ratio increases.Analysis employing PCA and SVD confirmed the existence of a set of robust modes that exist for the range of system parameters tested. Hence, modes shape obtained with different nozzles are the same for the same acceleration ratio of the jet flow. The existence of these structurally stable (in parameter space) fluid–structure modes has significant implications for the stability analysis of reactor fuel rods subjected to jet cross-flow. The results show that reduced order models of the high-dimensional fluid–structure can be developed and that the models can realistically capture the dynamics of the jet cross flow-induced vibrations of the fuel rod bundle.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.

CSNI status report “3D capabilities of thermalhydraulic system codes”: Lessons learned

In September 2015, an activity was endorsed by the OECD/CSNI/WGAMA to prepare a state of the art report presenting the 3D features of current thermal–hydraulic system codes. This activity was coordinated by IRSN and started in April 2016. This document is a synthesis of the full OECD report which is expected by 2023.Multi-dimensional phenomena that may have an important impact on the current thermalhydraulic models occur in the course of asymmetric scenarios such as feed water line break, main steam line break and loss of coolant accident. Most physical models were first established and validated in 1D geometry. The growing role of 3D modelling of reactor core and pressure vessel requires additional experimental data for a proper validation. Elements on available experimental support are proposed. However, the need of additional experimental support for validation has been identified in several directions such as the friction pressure losses, the diffusion–dispersion of momentum and energy dependency with space averaging, passive systems and the dispersion of void in core and downcomer. There are also several axes for progress, namely the development of phenomena identification and their ranking, the terms in the field equations and the mesh size dependence and the time step strategy.

A Petri net model-based resilience analysis of nuclear power plants under the threat of natural hazards

• Resilience of nuclear power plant against HILP events is assessed. • Petri net is effective in modelling the damage and maintenance of nuclear reactor. • The deterioration of the systems should not be neglected in the resilience study. • Periodic maintenance is critical for ensuring the resilience of nuclear reactor. Due to global climate change, nuclear power plants are increasingly exposed to the threats of extreme natural disasters. In this paper, a resilience engineering approach is applied to tackle all aspects of nuclear safety, spanning from design, operation, and maintenance to accident response and recovery, in the case of high-impact low-probability events. Petri net models are developed to simulate the losses caused by extreme events, the health states of relevant systems, mitigation processes, and the recovery and maintenance processes. The method developed is applied to assess the resilience of a single-unit pressurised heavy water reactor under the threat of three possible external events. Possible loss of coolant accidents and station blackout accidents caused by the events are considered. With the aid of the models developed, both the influence of stochastic deterioration and the impact of external events on the resilience of the reactor can be assessed quantitatively. The simulation results show that the method can comprehensively describe the resilience of nuclear power plants against various disruptive events. It is also found that the stochastic deterioration that does not directly affect the operation of nuclear reactors is critical to the resilience of reactors.

Research on simplification of branches method of accident sequences based on expert knowledge and heuristic optimization algorithm

The simplification of branches method of accident sequences based on expert knowledge and heuristic optimization algorithm is proposed in this paper, to increase the effectiveness of nuclear power plant safety margin analysis. First, the classification criteria for accident sequence and expert knowledge are used to screen out the deterministic failure sequences and deterministic safety sequences that are obviously. Secondly, a heuristic optimization method is used to study the variation in the accident sequences for those accident sequences that are challenging to evaluate with expert knowledge. Moreover, a hybrid particle swarm optimization (HPSO) is proposed to increase the accuracy of key safety parameters. This enables efficient accident sequence type identification, by comparing and evaluating key safety parameters and safety limits. Finally, the accident sequences are simplified using the cut-off frequency. To verify the effectiveness of the method, the small break loss of coolant (SBLOCA) accident in the nuclear power plant is analyzed in this paper. The numerical results show that the proposed HPSO improved the accuracy of solving the key safety parameters of accident sequences compared with the traditional method. By simplifying the accident sequences, the number of accident sequences requiring further analysis was reduced from 64 to 5.