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CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

The pressure control in the pressurized water reactor (PWR) primary loop is key to secure the safe operation. The pressurizer (PZR) unit is responsible for attaining this task. Thus, the PZR unit’s...

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

This research serves to advance the development of engineering computational fluid dynamics (CFD) computing efficiency for the analysis of pressurized water reactor (PWR) core using rod-type fuel assemblies with mixing vanes (one kind of typical PWR core). In this research, a CFD scheme based on the reconstruction of the initial fine flow field (RIF CFD scheme) is proposed and analyzed. The RIF scheme is based on the quantitative regulation of flow velocities in the rod-type PWR core and the principle that the CFD computing efficiency can be improved greatly by a perfect initialization. In this paper, it is discovered that the RIF scheme can significantly improve the computing efficiency of the CFD computation for the rod-type PWR core. Furthermore, the RIF scheme also can reduce the computing resources needed for effective data storage of the large fluid domain in a rod-type PWR core. Moreover, a flow-ranking RIF CFD scheme is also designed based on the ranking of the flow rate, which enhances the utilization of the flow field with a closed flow rate to reconstruct the fine flow field. The flow-ranking RIF CFD scheme also proved to be very effective in improving the CFD efficiency for the rod-type PWR core.

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

The lack of plant-side energy storage analysis to support nuclear power plants (NPP), has setup this research endeavor to understand the characteristics and role of specific storage technologies and the integration to an NPP. The paper provides a qualitative review of a wide range of configurations for integrating the energy storage system (ESS) to an operating NPP with pressurized water reactor (PWR). The role of ESS technologies most suitable for large-scale storage are evaluated, including thermal energy storage, compressed gas energy storage, and liquid air energy storage. The methods of integration to the NPP steam cycle are introduced and categorized as electrical, mechanical, and thermal, with a review on developments in the integration of ESS with an operating PWR. By adopting simplified off-design modeling for the steam turbines and heat exchangers, the results show the performance of the PWR steam cycle changes with respect to steam bypass rate for thermal and mechanical storage integration options. Analysis of the integrated system characteristics of proposed concepts for three different ESS suggests that certain storage technologies could support steady operation of an NPP. After having reviewed what have been accomplished through the years, the research team presents a list of possible future works.

This work presents an illustration of Nagra’s AMAC (Advanced Methodology for Activation Characterization) approach to the South Korean pressurized water reactor Kori-1 decommissioning. The results achieved are supported by the existing experience from the detailed AMAC applications to Swiss NPPs and are used not only for a demonstration of the applicability of AMAC to South Korean NPPs, but also for a first approximation of the activated waste volumes to be expected from Kori-1. A packaging concept based on the above activation characterization is also presented, using the AMAC algorithmic optimization software ALGOPACK leading to the minimum number of waste containers needed given the selected packaging constraints. Nagra’s AMAC enables effective planning before and during NPP decommissioning, including recommendations for cutting profiles for diverse reactor components and building structures. Finally, it is expected to lead to significant cost savings by reducing the number of expensive waste containers, by optimizing a potential melting strategy for metallic waste as well as by significantly limiting the number of radiological measurements. All information about Kori-1 used for the purpose of this study was collected from publicly available sources.

Inconel 690TT has an excellent high-temperature corrosion resistance and has been widely applied in steam generators of pressurized water reactor (PWR) nuclear power plants. In this paper, Inconel ...

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

Traditional light water reactor simulations are usually either high fidelity, requiring hundreds of node-hours, or low fidelity, requiring only seconds to run on a common workstation. In current re...

In nuclear power plants the design extension conditions are more complex and severe than those postulated as design basis accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6 inches (182 cm2) cold leg small break loss-of-coolant-accident to revise the effects of all safety systems ways through their availability/non-availability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of small break loss-of-coolant-accident as, Case-1, without scram and all of the safety systems are a failure, Case-2, the normal scram actuation with a failure of all safety systems (non-availability), and finally Case-3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creating a model using analysis of thermal-hydraulics of leaks and transient code. In all transient cases, all types of reactivity feed- backs, boron, moderator density, moderator temperature, and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literature. The results show the importance of the reactivity feedback effects in loss-of-coolant-accident on the fallouts power as they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal-hydraulic parameters are in acceptable ranges and encounter the safety criterion during loss-of-coolant-accident design extension conditions accidents processes. Furthermore, the results show that the core uncover and fuel heat up do not occur in KONVOI pressurized water reactor the design extension conditions simulations, as all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

Many experiments have been conducted on accidents and transients of pressurized water reactor (PWR) employing the rig of safety assessment/large-scale test facility (ROSA/LSTF). Recent research activities concerned with the OECD/NEA international joint projects included experimental investigation via the ROSA and ROSA-2 Projects, and counterpart testing with thermal-hydraulic integral test facilities under collaboration of the PKL-2, PKL-3, ATLAS, and ATLAS-2 Projects. Major results of the related integral effect tests (IETs) with the LSTF were reviewed to experimentally identify thermal-hydraulic phenomena involved, regarding the PWR accident sequences in accordance with the new regulatory requirements for the Japanese light-water nuclear power plants. Future separate effect test using the LSTF is planned to simulate loss of emergency core cooling system (ECCS) recirculation functions in a large-break loss-of-coolant accident (LOCA). Key results of the recent IETs utilizing the LSTF and future plans were presented relevant to multiple steam generator tube rupture accident with recovery operation, small-break LOCA with accident management measure on core exit temperature reliability, and small-break LOCA with thermal stratification under cold water injection from ECCS into cold legs. Also, main outcomes of the LSTF IETs were indicated for wide spectrum LOCA with core uncovery and anticipated transient without scram following small-break LOCA under totally failed high-pressure injection system.

Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed techniq...

We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.

Nuclear power plant, nuclear industry and aerospace industry require a high level of safety so it is necessary to make risk assessment for these systems, subsystems and components. So, it is needed to take into consideration the components failure rate and its probability, and have qualitative and quantitative evaluation using fault tree analysis. The conducted study is aiming to make qualitative and quantitative evaluation for Auxiliary Feed Water system (AFWS) which is designed to supply feed water to the steam generators for Reactor Coolant System (RCS) heat removal in a case of loss of main feed water or loss of all AC (alternating current) power to the station auxiliary, (AFWS) consisted of two trains to single steam generator which achieve 100% capacity for each train, Complete redundancy and diversity achieved. Results show the qualitative evaluation of auxiliary feed water system, this qualitative assessment is supported by failure rate data of components to have quantitative evaluation, this study indicated the unavailability of the individual components and the total unavailability of the (AFWS) for 8760 hour. To be sure that the system is available during operation periods, test and repair periods must be obtain at its exact scheduled time.

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer-, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of inter-assembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at AO Afrikantov OKBM in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of the core of VVER reactors.

China advanced large-scale Pressurized Water Reactor (PWR) implements series of passive safety systems to provide core cooling and decay heat removal in case of some unexpected accidents. The development of reactor best estimate system analysis codes has realized the transition from conservative safety analysis to Best Estimate plus Uncertainty (BEPU) analysis for the purpose of achieving more realistic results and digging larger safety margins. The BEPU analysis during Large Break Loss of Coolant Accident (LBLOCA) has been widely used for the licensing application of Nuclear Power Plants (NPPs). In this paper, the BEPU analysis of China advanced large-scale PWR was performed under the conditions of LBLOCA scenarios employing RELAP5 code. The models of reactor coolant system with all the passive safety systems were established and the influence of uncertainties of important parameters on the Peak Cladding Temperature (PCT) was evaluated. The results show the obtained PTC (1314 K) through the BEPU analysis is far below the 10CFR50.46 PCT acceptance criteria (1477 K) in case of LBLOCA and is also lower than the conservative calculation value (1348 K), indicating a larger safety margin. This work provides a meaningful evaluation method for China advanced large-scale PWR safety assessment.

Optimal control systems are crucial requirement that emphasize the safe operation of pressurized water reactor (PWR) nuclear power plants. The present study focuses on steam generator (SG) feedwate...

The economic assessment of advanced nuclear power reactors is very important, specifically during the early stages of concept design. Therefore, a study was performed to calculate the total cost estimation of fuel cycle supply for a system modular advanced reactor (SMART) by using the Generation-IV economic program called G4-ECONS (Generation 4 Excel-based Calculation of Nuclear Systems). In this study, the detailed description of each model and methodology are presented including facility, operations, construction matrix, post-production model, and fuel cycle cost estimation model. Based on these models, six Generation-III+ and Generation-IV nuclear reactors were simulated, namely System 80+ with benchmark data, System 80+ with uranium oxide (UOx) and mixed oxide (MOx) fuel assemblies, fast reactor, PBMR (Pebble Bed Modular Reactor), and PWR (Pressurized Water Reactor), with partially closed and benchmarked cases. The total levelized costs of these reactors were obtained, and it was observed that PBMR showed the lowest cost. The research was extended to work on the SMART reactor to calculate the total levelized fuel cycle cost, capital cost, capital component cost, fraction of capital spent, and sine curve spent pattern. To date, no work is being reported to calculate these parameters for the SMART reactor. It was observed that SMART is the most cost-effective reactor system among other proven and advanced pressurized water-based reactor systems. The main objective of the research is to verify and validate the G4-ECONS model to be used for other innovative nuclear reactors.

Samll size and highly integral micro nuclear reactors which have broad energy advantages in the application of ocean, land, space, and sky, become a hot research topic in the international nuclear industry recently. Compared with a large-scale pressurized water reactor nuclear plant composed of a large amount of subsystems, the integral micro nuclear reactor has fewer subsystems but its subsystems are tightly coupled due to the constraints of the volume and weight. To handle the complexity aroused from this coupling issue and to better predict the reactor dynamic behavior, it is necessary to perform the system simulation across multiple disciplines and domains. In this paper, Modelica language was used in the system modeling and simulation of a micro gas-cooled reactor. The Modelica model was self-developed by China Nuclear Power Engineering Company and the MWorks developed by Suzhou TongYuan was chosen as the simulation platform. Two simulations of a conceptual micro gas-cooled reactor design were carried out, including an extreme accident scenario and a normal load-following operation scenario. The simulation results showed that the reactor has good inherent safety even under the extreme accident, in which the reactor shutdown can be achieved only by the negative reactivity result from the increase of core temperature and the fuels were not damaged since the decay heat was removed by passive air cooling from outside of the reactor pressure vessel. The reactor also has good load-following performance, which can be achieved by simply adjusting the helium inventory (or pressure) and the control rod position, while the core temperature and power generation efficiency kept constant.

Multiple spurious operations (MSOs) mean multiple fire induced circuit faults causing an undesired operation of one or more systems or components. The Nuclear Energy Institute (NEI) of the United States published NEI 00-01 as guidelines for solving MSOs. And this guideline includes MSO scenarios of pressurized water reactor (PWR) and boiling water reactor (BWR). Nuclear power plant operators in U.S. analyzed MSOs under MSO scenarios included in NEI 00-01 and operators of PWRs in Korea also analyzed MSOs under the scenarios of NEI 00-01. As there are no pressurized heavy water reactors (PHWRs) in the United States, MSO scenarios of PHWRs are not included in the NEI 00-01 and any feasible scenarios have not been developed. This paper developed MSO scenarios which can be applied to PHWRs by reviewing the 63 MSO scenarios included in NEI 00-01. This study found that seven scenarios out of the 63 MSO scenarios can be applied and three more scenarios need to be developed. Keywords: Fire hazard analysis (FHA), Fire safe shutdown analysis(FSSA), Multiple spurious operation (MSO)

The steel liner dome of containment structures has a large span and small thickness, and is attached to the concrete with welded studs and angle steel section anchorages. It acts as the interior formwork during pouring of the concrete, but there is no support or brace for the steel liner dome owing to the construction difficulty, which makes the issue of stability during construction very important. A pressurized water reactor containment was used as an example to investigate the stability of a steel liner dome during construction. In the first key construction phase, stability analysis of the steel liner dome was carried out systematically to obtain the buckling modes and plasticity development. The effect of initial geometrical imperfections on the stability of the steel liner dome was investigated, and a prediction equation for the stability bearing capacity was proposed. In the second key construction phase, the functional failure of the steel liner–concrete composite dome was investigated, and the effects on the buckling loads of composite domes were examined. The results show that the transition area of the dome at which the radius of curvature changes is the weak area in the first key construction phase. Dents occur at this transition area, and the plasticity development at the transition area results in the buckling of the whole structure. The effect of initial geometrical imperfections on the ultimate bearing capacity is not significant. The functional failure is cracking in the ring beam during the second key construction phase.

The fuel rod performance has great importance for the safety and economy of an operating reactor. The fuel rod performance analysis code, which considers the thermal-mechanical response and irradiation effects of fuel rod, is usually developed in order to predict fuel rod performance accurately. The FRIPAC (Fuel Rod Integral Performance Analysis Code) is such a fuel rod performance analysis code that has been developed recently by China Nuclear Power Technology Research Institute Co. Ltd. The code aims at the computational simulation of the Pressurized Water Reactor fuel rod behavior for both steady-state and power ramp condition. A brief overview of FRIPAC is presented including the computational framework and the main behavioral models. Validation of the code is also presented and it focuses on the fuel rod behavior including fuel center temperature, fission gas release, rod internal pressure/internal void volume, cladding outer diameter and cladding corrosion thickness. The validation is based on experimental data from several international projects. The validation results indicate that FRIPAC is an accurate and reliable fuel rod performance analysis code because of the satisfactory comparison results between the experimental measurements and the code predictions. Keywords: FRIPAC, Fuel behavior, Model, Validation

The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method.

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics. Keywords: Fluid-structure interaction, U-tube, Steam generator, CFD, Flow-induced vibration

Operation with reduced number of operators is considered as an important operational mode to improve economics of a nuclear power plant (NPP) in the future and it will be more important for the small modular reactors (SMRs). To control NPPs efficiently and safely with a reduced number of operators, an intelligent operation system can be very useful. However, reduced number of operators will result in larger consequence in human error due to simultaneous multi-unit failures are now more probable than the conventional operational mode. Thus, to decrease the probability of human error occurrence for the reduced number of operators, the development of an autonomous operation system with certain level of intelligence is inevitable. The prediction capability of system’s behavior as well as understanding the root cause for the current situation should be an essential function of the designed autonomous operation system. The authors propose to use the current system thermal–hydraulic (STH) analysis code platform to develop such intelligence. As a demonstration of the suggested method, the authors used the STH code to track the accident initiator of both large pressurized water reactor (PWR) and SMR. The discussions on how this method can be useful for future NPP operation and how generic differences in system design actually influence the system behavior are presented.

Small modular reactor (SMR) is a type of advanced reactor depending on its inherent safety, good economic performance, and multipurpose. SMR always adopts the multi-modular layout, which uses several nuclear steam supply systems (NSSSs) to supply steam to a common steam header. A pressurized water reactor (PWR) type multi-modular SMR (MSMR) system is studied in this paper. Due to the operation demands of continuity and flexibility during load following, a proper control strategy and operation scheme of the MSMR must be investigated. A master-slave control strategy is proposed for steam flow sharing control of the MSMR systems. Under the proposed control strategy, one of the NSSS modules is selected as the master control channel, and its feed water flow rate is regulated using the steam header pressure error. The remaining NSSS modules are treated as the slave channels. The feed water flow rate of the master channel is utilized to slave the other channels, which are flow-controlled. Three operation schemes are set according to different operation demands. (1) Equilibrium Scheme (EquS), in which every module undertakes the same load during the load following operation. (2) Non-equilibrium Scheme (NEquS), in which every module undertakes the different loads during the load following operation. This operation scheme manages more modules operate at full power to ensure its economy and operation simplicity. (3) Constant Output Scheme (ConS), which is mainly used during refueling or maintenance process. When one module is disturbed, other modules undertake its load to ensure the total power output constant. Based on these schemes, four typical operation cases are defined and simulated. Through the analysis of the simulation results, the suitable operation conditions for these operation schemes are suggested, and the relevant control strategy is verified.

This work proposed a new method for prediction of hydrogen Deflagration to Detonation Transition (DDT) on the basis of oxygen concentration in the presence of inerting diluents. Whereas previously, the traditional criterion for deflagration to detonation transition hypothesized an unchanged air composition, it now seems appropriate to question the assumption and consider possible situations in which the presence of inerting gas components incapacitates the old criterion for applications. Under some circumstances (severe accidents in nuclear power plants), hydrogen may be massively generated by intense chemical reactions between zirconium cladding and overheated coolant in the nuclear reactor vessel. In order to prevent hydrogen explosions, Passive Autocatalytic Recombiners (PARs) that mitigate hydrogen risk by hydrogen oxidations have been implemented in the nuclear energy industry worldwide. It consumes a large amount of oxygen as the reactant and gives rise to an increased ratio of inert gas nitrogen to oxygen in the air, the product of which, water mist, also alleviates explosion hazards. The new method addressed on the variation of oxidant volume fraction and proposed new parameters: the equivalent air and the equivalent inert gases concentrations in deflagration to detonation transition criterion. The HYDRAGON code, that has been specially developed for hydrogen analysis in nuclear power plants, implemented both new and original criteria and has been applied to assessments. Close agreements between numerical simulations and a large number of experimental data sets: a wide variety of fuel gases and inert diluents, suggested that such new technique was viable and applicable to predict deflagration to detonation transition for various combustible gases. A hydrogen risk analysis of an advanced pressurized water reactor using the new method was also demonstrated in this paper.

Nuclear fission reaction is a clean energy source that can provide thermal power to desalination processes for producing potable water. Due to the inherent safety features of the integral pressurized water reactor (iPWRs) and economic competitiveness of the multi-effect-desalination and thermal-vapor-compression (MED-TVC) process, it is reasonable to interconnect the iPWR with MED-TVC process to develop nuclear desalination plants (NDPs). Dynamic modeling is important for both the design and optimization of the control systems for nuclear desalination plants. In this paper, based on the balance of mass, momentum, energy and salt, a lumped-parameter dynamic model for the NDP composed of an iPWR named NHR-200II and the MED-TVC process is given for the plant control design, which is further deployed as a program on Matlab/Simulink platform. For the modeling verification, the steady points of some key process variables are shown with comparison to the design data, and the open-loop responses under the stepping of both the input seawater flowrate and TVC throat cross-section area are given, which not only show accuracy of this model but also illustrate the coupling between the nuclear steam supply system (NSSS) and MED-TVC modules. Then, a plant control strategy of the NDP based upon NHR-200II reactor and MED-TVC module is newly proposed. Based on the dynamic model, the control verification is finally performed, and the closed-loop responses in the cases of both power maneuvering and MED-TVC module bypassing show the feasibility and satisfactory performance of this control design.

The most important public concerns is the safety of nuclear power reactors. It is extremely important to protect the public from the nuclear radiation exposure in the undesired case of a nuclear power plant (NPP) accident. The proper performance of the safety systems of nuclear power plants is one of the most highly important concerns which enhances the use of nuclear energy options. This research paper focuses on the response and performance of the emergency safety systems of a pressurized water reactor (PWR) during loss of coolant accident (LOCA) for a selected reference PWR. The simulation depicts the reactor behavior during hot leg loss of coolant accident event and the leakage area is 200 cm2. This was accomplished by modeling the PWR and the passive safety systems employed using PCTran code which is designed to describe the emergency core cooling system (ECCS) response and core behavior under severe accident conditions. The ECCSs are designed to cool down the reactor core and to prevent core damage. Results of the core temperature distribution, reactor coolant inlet and outlet temperatures, reactor coolant mass flow rate, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the PCTran model is capable of contributing to the process of safety analysis of the NPP.

The nuclear accident identification problem has been developed quite extensively in the literature, with numerous statistical methods and artificial intelligence techniques having been employed over the years to deal with this important safety matter. This article presents, for the first time, the application of genetic programming to this problem. The methodology consisted in evaluating the efficiency of the algorithm as a technique for the optimization and feature generation in a pattern recognition system for the diagnostic of accidents in a pressurized water reactor nuclear power plant. Considering the set of the time evolution of four physical variables for the three accident scenarios approached, plus normal condition, the task of genetic programming was to evolve non-linear classifiers that would provide the largest amount of discriminatory information for each of the events and, consequently, better identification rates. Genetic programming was proven to be a methodology capable of attaining success rates of, or very close to, 100%, with quite simple parameterization of the algorithm and at a very reasonable time, putting itself in levels of performance similar or even superior to other similar systems available in the scientific literature, while also having the additional advantage of requiring very little pretreatment, or a priori knowledge, of the data.

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.

In an iPWR (integral pressurized water reactor), the pressurizer is located at the top of the reactor vessel; this configuration requires the investigation of the mechanisms adopted for the homogenization of boron. In previous work, three representative experiments were conducted in an experimental installation made of stainless steel with a volumetric scale equal to 1:200, representing one-fourth of the pressurizer of an iPWR. The test facility layout was mounted so that the test section was fed with water or saline solutions with different concentrations, representing scenarios of boration or deboration. To determine the concentration at the inlet and outlet of the test section, samples have been collected every 10 min during the experiments representing some scenarios. The main goal in this work was to investigate how well mixing during boron dilution transients in the pressurizer of a small modular reactor can be modelled accurately by CFD (Computational Fluid Dynamics) codes. Two CFD codes were used, the commercial ANSYS CFX and the open source CFD code OpenFOAM. The use of open source software such as OpenFOAM offers a way to CFD gain acceptance in the licensing. The results of the comparison of simulations with an experiment at the test section were presented and showed a very good agreement. It was verified that deviations are less than 4%, both codes can be used to accurately represent this phenomenon. In order to improve future experiments, the dispersion of the solution inside the test section is studied with the aim to propose positions for sensors. A first simple configuration is proposed, but some further simulations will be done to find an optimum configuration.

The design of Instrumentation and Control (I&C) system in nuclear power plan needs to consider the problems of Common Cause Failure (CCF) and independence of itself in addition to meeting the overall Defence in Depth (DiD) target of the plant and ensuring the correct and reliable implementation of monitoring, control and protection functions under various operation conditions. The requirements of International Atomic Energy Agency (IAEA) are summarized in this paper. A design scheme of DiD for I&C system of Nuclear Power Plant (NPP) is introduced. in this paper. The compliance of DiD design for I&C system with the requirements of IAEA is analyzed. The analysis concludes that the I&C design scheme basically meets the requirements of IAEA. At the same time, the improvement of diversity should be further studied. The study of this paper provides valuable reference for the continuous improvement of the design of I&C systems in NPP.

Digital instrumentation and control (I&C) system has been widely applied on large commercial pressurized water reactors (PWR), which brings significant convenience to their operation. However, the level of automation and operation efficiency of nuclear power plant is still lower than conventional power plant such as thermal power plant or hydro power plant. On one hand, the safety and reliability requirement of nuclear power plant always chooses the mature and proved technology, but not the emerging technology such as artificial intelligence. On the other hand, the complexity of system design and operation requirement also makes it difficult to improve the automation level of nuclear power plant. To deal with these difficulties, a statistical analysis has been given to identify the challenges on improving the automation level. And a framework of plant level autonomous is given to illustrate how to organize the techniques and algorithms available to reach this goal. For the new functions, algorithms and equipments that involved in the framework, the requirement of nuclear safety should be considered. Therefore related requirements that need to be addressed are also discussed in this paper.

Advanced manufacturing has the potential to revitalize US manufacturing, with valuable applications in several industries, including aerospace, automotive, and construction. Some of these applications have clear-cut objectives (e.g., maintain component performance while reducing mass). Applications of advanced manufacturing of nuclear components have aimed at recapturing lost manufacturing capabilities or addressing maintenance of legacy reactor components. Through the Department of Energy, Office of Nuclear Energy, Transformational Challenge Reactor design and analysis thrust, applications of advanced manufacturing, in particular, additive manufacturing, to core design has yielded reactor designs that are free from conventional manufacturing constraints. For applications in core design, the multiphysics nature of the key core metrics (e.g., peak temperature, peak power) in addition to transient safety performance requirements provides a more complex set of objectives that requires more advanced modeling and simulation tools. Additive manufacturing provides high dimensional control and design flexibility to produce complex coolant channel shapes for improved heat transfer properties and low peak material temperatures. Additional mechanisms for improved heat transfer characteristics and temperature-controlled feedback mechanisms have also been explored and incorporated into designs. While some of these enhancements are not directly beneficial for the current operating pressurized water reactor fleet, benefits may be realized in specific reactor applications that have a more constrained design space (e.g., mass, size, material type) or design metrics (e.g., fuel utilization).

For pressurized water reactor nuclear power plants, in order to prevent the release of radioactive substances into environment, fission product barriers (FPBs) are constructed based on the concept of defense-in-depth, including fuel clad, reactor coolant system (RCS), and containment; the status of these FPBs is then acting as an important dimension to decision-making of emergency action levels (EALs). For CPR1000 nuclear power plants, state functions defined in state-oriented emergency operating procedure (SOP) are used to characterize postaccident physical conditions; their degradation substantially represents the challenges on fundamental safety functions and then on the integrity of FPBs in like manner, so degradation of these state functions is referred to as determining initial conditions of each FPB, by which the link between SOP and EALs is established. Then, an intelligent FPB monitoring system (FPBMS) aiming to automatically monitor states of FPBs is developed, verified, and validated. The pioneering work, by building bridges between state functions and initial conditions of FPBs and then computerizing them innovatively, proves that dynamical monitoring of states of FPBs during accident evolvement and real-time indication of loss or potential loss of FPBs can be achieved, which is most helpful in decision-making of EALs.

International audience; Identifying what causes fuel assembly vibrations downstream of Mixing Vane Grids (MVG) in Pressurized Water reactor (PWR) is of paramount importance for nuclear community to understand grid-to-rod fretting wear. Experiments, called CALIFS, were carried out by the Atomic Energy Commission (CEA) on a 5x5 MVG at a hydraulic Reynolds number of 66,000, in order to measure the flow velocity and the pressure along the central rod. In parallel, a benchmark for Large Eddy Simulation (LES) was setup to compare the predictions of three different CFD codes: Star-CCM+, Code_Saturne and TrioCFD to experimental measurements. The computational domain is representative of a span of CALIFS mockup, composed of a 5x5 rod bundle with a MVG. The three computations overall give very satisfactory results, independently from the mesh created and the modelling options selected. It seems to suggest that whatever the software used, this kind of calculations has reached a significant level of robustness and accuracy. Nonetheless, some discrepancies remain concerning the predictions of pressure standard deviation decay far downstream of the mixing vane grid.

The key technologies such as accident monitoring, installation and positioning, and signal abnormality of Nuclear Instrumentation System (hereinafter referred to as RPN) is studied in this paper. Through analyzing the requirements of post-accident monitoring laws and regulations, the compensation ionization chamber monitoring technology and fission chamber monitoring technology are deeply studied. The positioning mechanism of detector components is analyzed, and the positioning technical characteristics of “push-pull trolley type” and “bucket type” are systematically studied. The signal abnormality phenomenon is analyzed, and the equivalent circuit diagram and fault tree are established to analyze the technical causes of signal abnormality. Through the research on the key technologies, an advanced system scheme is independently realized. The independent scheme adopts fission chamber accident monitoring technology, realizes “bucket type” installation of detector components, and adopts measures such as improving gateway redundancy to ensure reliable and safe operation of the unit. The autonomous scheme has been applied to an advanced third-generation pressurized water reactor in China, and the research results have certain reference value for the research and design improvement of other reactors.

Rising awareness for the environment as well as concerns over the sustainability of fossil fuels has encouraged developed and developing countries to find alternative ways to enhance the thermal efficiency of current power systems. The thermal efficiency of power plants can be increased from 30 – 40 % up to 80 – 90 % through the implementation of a trigeneration system by recovering dissipated waste heat for other purposes. The trigeneration system can be defined as a technology that can produce simultaneous power, heating, and cooling energy from the same fuel source. Trigeneration System Cascade Analysis (TriGenSCA) methodology is an optimisation approach based on Pinch Analysis that has been used to establish the guidelines or the proper size of the trigeneration system. This paper proposes a modification of TriGenSCA by considering a multi-period of energy consumption to optimise the size of the utility in the centralised trigeneration system by considering the transmission and storage of energy losses in the Total Site system. There are six steps involved including data extraction, identification of time slices, Problem Table Algorithm (PTA), Multiple Utility Problem Table Algorithm (MU PTA), Total Site Problem Table Algorithm (TS PTA), and modified TriGenSCA. The methodology has been tested on the centralised nuclear trigeneration system in a Total Site System as a case study and results shown that thermal energy needed by the Pressurized Water Reactor (PWR) trigeneration system with transmission losses is 2,427 MW whereas thermal energy needed by the PWR trigeneration system without transmission losses is 2,424 MW. The TriGenSCA with consideration of transmission and storage energy losses is useful for engineers and designers to determine the exact value of energy for trigeneration plant.

This paper takes verification of the reactor RRC system of a pressurized water reactor nuclear power plant as the research object. This paper determines the modelling scope of the RRC system according to the characteristics of the RRC system, modeled its process systems and control systems, and eventually constructed the digital Twins system of the RRC system, which realizes the dynamic closed-loop verification of the RRC system. The paper gives examples of implementation for typical accidents. The research results eventually achieve the purpose of accurate verification of the RRC system, and improve the safety and economy level of nuclear power plants, and play an important role in improving the safety and economy level of nuclear power plant operation.

SMR has its easy receptivity of advanced design concepts and technology. Inherent safety features and passive safety systems realize drastic safety enhancement, and system simplification and modularization can maximize its economics. As they successfully completed the technology self-reliance program for the design and construction of pressurized water reactor (PWR) plant with American companies, Korea has pursued on technology development for SMR (small modular reactors) for the last two decades. This chapter introduces their various SMR R&D programs with their design characteristics and future prospect. Representatively, the Korea Atomic Energy Research Institute (KAERI) has developed integral type PWR—i.e., SMART to acquire the standard design approval according to Korean regulatory frame, and is making efforts to improve safety further. Sodium-cooled fast reactor (SFR) and very high temperature gas-cooled reactor (HTGR) designs are also being developed in KAERI. Besides KAERI’s research, several developments for the various types of SMR’s have also been made by the other Korean organizations such as KEPCO E&C, Seoul National University (SNU), KAIST, and UNIST, whose reactors are different in size, power, core, and application.

A nuclear submarine has significant performance advantages over the conventional diesel-electric submarine. Its powerful nuclear propulsion can be operated independently of air and does not require refueling throughout the 25-year life span. Recent progresses of medium voltage direct current (MVDC) techniques applied in industry push the energy-saving and extremely long-time operation to a new stage with a feasible solution of smaller scale nuclear reactor embedded inside the vessel. Thus, there is a need to investigate the new MVDC-based nuclear submarine power distribution system and its real-time hardware-in-the-loop (HIL) emulation to test the control systems, device stresses, etc. This paper proposes a comprehensive MVDC-based nuclear submarine power distribution system emulated in real-time on a multi-FPGA hardware platform with system-level transients (validated by Matlab/Simulink) at 20 microsecondtime-step and device-level transients (validated by SaberRD) at 400 nanosecondtime-step.

Design-by-analysis (DBA) procedures for Nuclear Class 1 pressure vessels such as those prescribed within ASME Boiler and Pressure Vessel Code (BPVC) Section III, provide rules to demonstrate assurance against fatigue failure. Two general assessment routes exist, linear finite element analysis (FEA) with stress categorization and elastic-plastic penalty factors, or nonlinear FEA with direct multiaxial strain evaluation. Whilst the elastic design route possesses many practical advantages, it is widely acknowledged to be very conservative, sometimes unacceptably so. At the cost of additional analysis effort, plastic design methods can provide a more appropriate evaluation of fatigue usage, potentially avoiding unnecessary design modifications and reducing the burden of in-service inspection requirements. This paper presents and compares various strain measures proposed for ASME III plastic fatigue analysis within the technical literature. A case study of a typical pressurized water reactor (PWR) main coolant line (MCL) piping nozzle subjected to pressure and thermal loads is presented. The influence of strain measure selection on the FE-derived strain concentration (Ke) factors is examined. Some important considerations for calculation of realistic Ke factors in ASME III are further discussed.

Power grid fault will cause the frequency and voltage fluctuation of the nuclear power plant, which will directly affect the operation status of the coolant pump and the heat dissipation blade of the nuclear power unit. When the fluctuation degree exceeds the maximum allowable range of nuclear power equipment, may cause the nuclear power plant damage. In order to study the operation status of large-scale nuclear power plant equipment under grid failures, this article focuses on nuclear power plants in a certain area. A 500 kV grid model including pressurized water reactor(PWR) nuclear power plants is built on the PSCAD/EMTDC platform. The custom details of nuclear power plants are detailed The model includes reactor core, steam generator, steam turbine module, etc. The model can reflect the internal dynamic characteristics of nuclear power plant equipment in the event of a grid failure. By simulating the impact of voltage and frequency disturbances on nuclear power plants in the case of grid faults, the disturbance tolerance threshold for the safe operation of nuclear power plants is given, and equipment protection measures in the event of a three-phase short-circuit fault in the power grid are proposed, which is the fault ride-through of nuclear power plants. Further research on capabilities has laid the foundation.