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AbstractIn this paper, a dynamic prediction scheme that combines the data assimilation method and dynamic mode decomposition (DMD) is brought out for the prediction of the whole-core power distribution under xenon oscillations within the HRP1000 reactor. The DMD is used to predict the power values over the nodes where in-core detectors exist, and predicted power is then extended to the whole core using data assimilation methodologies, e.g. the inverse distance–based data assimilation method. In the data assimilation stage, the selection of the background physical field and the regularization factor under different noise levels is investigated. A series of numerical experiments, based on the HPR1000 proof of feasibility of the coupling scheme, is conducted under low noise levels or low prediction step sizes. Finally, the optimal application conditions and the prediction performance of the coupling scheme in different noise levels are analyzed for practical engineering usage. [ABSTRACT FROM AUTHOR]

Difference from the movable neutron detector for CPR1000 reactor can only be used periodically, HPR1000 and other Gen-Ⅲ reactor innovatively use self-powered neutron detector (SPND) to continuously measure the in-core neutron flux. This paper provided detailed design and implementation of the high-precision core surveillance system for the first HPR1000 reactor. Firstly, a novel SPND signal processing method was proposed, which not only solved the limitation of traditional design software for high-precision SPND current calculation, but also enabled high-precision signal processing of measured SPND. Based on this, a high-precision core 3D power reconstruction method was established. Subsequently, the online monitoring system SOPHORA for the HPR1000 reactor core designed based on this theoretical model was described, and its uncertainty analysis method was explained in detail. The uncertainty analysis results indicate that the accuracy of the key parameters of the system can meet the needs of high-precision core monitoring. Finally, in order to confirm the performance of the system, a comparative analysis was conducted on the deviation between the measurement and theoretical prediction of SPND current during the start-up process of the HPR1000 reactor, as well as the deviation of component power distribution. The results show that the deviation is much smaller than the regulatory requirements for initiating physical tests. The double validation of the uncertainty analysis results and the component power deviation analysis results during the start-up process indicates that SOPHORA achieves high-precision core monitoring, and its use of SPND signal processing and core 3D power reconstruction methods has significant accuracy and reliability.

The reactor operation digital twin provides real-time parameter and state estimation for the reactor during operation, providing input for subsequent safety parameter calculations. As one of the core modules of the reactor operation digital twin, the inverse problem solver is crucial to ensure the real-time and accuracy of parameter and state estimation. In the previous work, the inverse problem is solved based on an initial guess of the input parameter, and the quality is highly depend on the initial guess. In order to improve the accuracy and computational efficiency, a global-local search (GLS) method was proposed in this work. Firstly, the reduced order method and KNN (K-nearest neighbor) method were used to build a reduced forward model, with which one can compute the physical field for a given input parameter online. The inverse problem, i.e., the process of calculating from a set of observations the causal factors that produced them, was then solved using GLS. In the global search stage, the KNN method was used to predict an initial input parameter based on the observed values. In the local search stage, the Latin hypercube sampling (LHS) method was used to discrete the local neighborhood of the initial parameters. The theoretical observation values corresponding to the parameters were calculated based on the reduced forward model, which was also trained using KNN. The optimal input parameter was then determined with which the observations were closest to the actual observation values. Numerical tests were conducted on HPR1000 reactor operation digital twin. To simulate the power distribution of HPR1000, the burnup, the power rate, the control rod inserting steps and the inlet temperature of the coolant were selected as input parameters, and the neutronic code CORCA-3D was used to calculate the power distribution with the four-dimensional input parameters. The digital twin was constructed over a wide range of the four-dimensional input parameters. The synthetic observations, calculated with CORCA-3D, were used to simulate the real observations of the in-core self-powered neutron detectors (SPNDs). The inverse problem solvers with different methods were used to infer the four-dimensional input parameters using the synthetic observations. The accuracy and computational efficiency for parameter inference and the related physical field with and without noisy observations were investigated using GLS, KNN and LHS. Numerical results verify that the proposed GLS outperforms other methods, and provides real-time and accurate parameter and state estimation for reactor operation digital twin, laying the foundation for engineering practice.

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Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Introduction: The accurate prognosis of reactor accidents is essential for deploying effective strategies that prevent radioactive releases. However, research in the nuclear sector is limited. This paper introduces a novel Temporal Fusion Transformer (TFT) model-based method for accident prognosis that incorporates multi-headed self-attention and gating mechanisms. Methods: Our proposed method combines multi-headed self-attention and gating mechanisms of TFT with multiple covariates to enhance prediction accuracy. Additionally, we employ quantile regression for uncertainty assessment. We apply this method to the HPR1000 reactor to predict outcomes following loss of coolant accidents (LOCAs). Results: The experimental results reveal that our proposed method outperforms existing deep learning-based prediction models in both prediction accuracy and confidence intervals. We also demonstrate increased robustness through interference experiments with varying signal-to-noise ratios and ablation studies on static covariates. Discussion: Our method contributes to the development of intelligent and reduced-staff maintenance methods for reactor systems, showcasing its ability to effectively extract and utilize features of static and historical covariates for improved predictive performance. [ABSTRACT FROM AUTHOR]

Copyright of Journal of Drainage & Irrigation Machinery Engineering / Paiguan Jixie Gongcheng Xuebao is the property of Editorial Department of Drainage & Irrigation Machinery Engineering and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Under severe accident conditions, if the coating in the containment peels off, it may lead to blockage of in-containment refueling water storage tank (IRWST) filters, which cannot ensure the availability of the core injection. Therefore, the performance of the coating used in the containment of HPR1000 should be considered in the selection of the coating to ensure the normal operation of IRWST filters under the serious accident conditions. According to HPR1000 serious accident occurred inside the containment of coating environment conditions, the innovative design is given in terms of the simulation test for coating process in severe accident, the test project and acceptance requirements. In addition, coating products with better resistance to radiation and high temperature resistant performance are selected for verification. The appearance and adhesion of the tested samples are also tested. The test results show that the coating after serious accident condition simulation test can satisfy engineering demand. It ensures that a serious accident probably happened in HPR1000 won't produce paint slag crushing and block IRWST filters, thus ensures the availability of the core injection and then the corium melt to be retained in the reactor, and further enhance the response ability for severe accident of HPR1000. [ABSTRACT FROM AUTHOR]

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m2. The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20–50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

Extensive studies have been carried out on the behavior of core degradation and fission products of common pressurized water reactors (PWRs). However, few of them have investigated the relationship between thermal hydraulic and fission product behavior in advanced passive PWRs. Due to the impact of thermal hydraulic behaviors in different accident sequences on the release and transportation of fission products, an integrated severe accident analysis (ISAA) code with highly coupled thermal hydraulic and source term calculations is required to simultaneously analyze thermal hydraulic and source term behavior. For advanced passive PWRs, important safety systems that may affect the behavior of the core and fission products should be considered. It is therefore necessary to simulate the thermal hydraulic and fission product behavior of advanced passive PWRs. In this study, the ISAA code is adopted to simulate the occurrence of a hypothetical double ended cold leg LBLOCA of HPR1000 in three scenarios of equipment failure. The results show that the high-temperature fuel rods and cladding materials exhibit delayed failure at the lower position of the active core, whereas earlier failure at higher position during the reflooding. Active and passive equipment affects fuel temperature, the oxidation conditions of the fuel, the interaction of fission products and structural materials, and the state of the fuel, thereby affecting the release of fission products in the fuel. HPR1000 only relies on passive equipment to relieve the core degradation in severe accidents, realize the in-vessel retention of melt, and eliminate the ex-vessel release possibility of fission product. It is hoped that the results can provide references for HPR1000 to formulate the severe accident management guidelines (SAMG).

Nuclear power plays a crucial role in achieving the target of carbon neutrality to build a sustainable society. However, it is not “carbon-free” when considering its entire life cycle. Therefore, accurate accounting and monitoring of its generated carbon emissions are required to avoid miscalculations of nuclear energy as a clean energy source. In this study, the life-cycle carbon emissions of nuclear power plants (NPPs) with different reactor types are reviewed. In addition to the characteristic differences among different reactors, disparities in the review results originate from the varying emissions at the respective stages of the nuclear fuel cycle, technology choices at each stage and accounting methods and boundaries. The carbon emissions resulting from NPP construction and operation are underestimated due to the limited data and methods, which creates uncertainty in the evaluation of NPP carbon emissions. An integrated framework for carbon emissions accounting considering the construction and operation of NPPs (CACO-NPP) is proposed. This integrated framework aims to improve the accounting accuracy for carbon emissions originating from NPPs. An emerging Generation III NPP with the latest technology, HPR1000 (an advanced pressurized water reactor), was adopted as a case study. The results show that the total emissions resulting from vegetation loss, equipment manufacturing and labor input during construction and operation are 1232.91 Gg CO2 with a carbon intensity of 1.31 g CO2/kWh, indicating the notable mitigation capability of Generation III NPPs. By combining the maturity of HPR1000 technology with successive design improvements, the carbon emissions of such reactor types could be further reduced. This development is very important for realizing China’s carbon neutrality target.

[Introduction] The pressure transmitter important to safety of nuclear power plant needs to pass the equipment qualification to confirm that it can perform the safety function under the accident condition. However, the current relevant standards do not explain the logic and mechanism of such equipment qualification, the rationality is not sufficient. [Methods] Based on the understanding of the identification system of electrical equipment important to safety at home and abroad, combined with the actual identification process of pressure transmitter important to safety of HPR1000, the identification theory of pressure transmitter important to safety was studied in detail. [Results] The important operating environment factors affecting its safety performance were analyzed, the significant aging mechanism was identified, the accelerated aging test was described based on the accelerated aging mathematical model, and the typical accelerated aging sequence and identification process of the safety-important pressure transmitter of HPR1000 was recommended. [Conclusion] The identification process can be used as a reference for the identification of safety-important pressure transmitters in China.

• The state space model of NSSS of HPR1000 is established. • A state feedback controller based on the dominant pole is designed. • A new coolant average temperature control system is designed. • The particle swarm optimization (PSO) is used to optimize the relevant parameters. • The control system designed in this paper has better control performance. Control of nuclear power plant is still based on the traditional PID control system, which is difficult to obtain high control quality in the process of a wide range of load changes. To effectively use the measurable information of the system and consider the constraints, state feedback control based on the dominant pole method is proposed for the average coolant temperature control of HPR1000. The control system is divided into two parts: one part is a feedback branch, which realizes the state feedback by using the measurable system state quantity including the core inlet temperature, the core outlet temperature and the reactor power, and at the same time introduces the integral link to reduce the steady-state error; the other part is a feedforward branch, which uses the nominal load change to make feedforward compensation to improve the control performance of load tracking. At the same time, Particle Swarm Optimization (PSO) method is used to optimize the controller parameters, and the dominant pole meeting the requirements is obtained. The control performance under different working conditions is verified on the HPR1000 model. The test results show that the state feedback control can effectively improve the setpoint tracking ability and anti-disturbance ability. [ABSTRACT FROM AUTHOR]

HPR1000 NPP is designed with Passive Containment Heat Removal System (PCS) following the technical route called active and passive combined concept. The installed PCS system will have effect on the thermal hydraulic (T-H) behavior in the containment, and the complex T-H environment will influence the heat removal and operating character of PCS system backwards. The main goal of this work is to evaluate the effectiveness of the PCS system of HPR1000 and analyse the complex T-H environment of HPR1000 when the PCS system is operating. The HPR1000 containment is modelled by the containment T-H code and the PCS system is modelled based on experiment results. Then, the severe accident sequence initiated by large break LOCA is calculated by an integrated computer code, which outputs mass and energy release sources from reactor coolant primary system as boundaries for the containment T-H code. The results show that HPR1000 PCS system has enough heat removal capability under selected severe accident, which could ensure the containment meet the design requirement. Also, the nonuniformity of temperature and gases in the containment will disappear when the mass and energy releases turn stable, and the operating of PCS system does not cause a huge effect on containment inhomogeneity.

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Event sequences that would lead to an early radioactive release or a large radioactive release are required to be practically eliminated in HAF102—2016. But there are lack of specific acceptance criteria and demonstration method of practical elimination in China. Safety requirement of practical elimination was studied, and some insights, acceptance criteria and demonstration method of practical elimination were proposed in this paper. Demonstration of practical elimination for HPR1000 was evaluated. The main conclusions are as follows: 1) Practical elimination is a higher requirement for the safety design of nuclear power plants. It is not a requirement that accident conditions have no release, but the plant states that could lead to an early or a large radioactive release have been practically eliminated. And conditions that have not been practically eliminated should be fully considered in the design to ensure its radiological consequences limited. 2) With reference to international practice and the relevant requirements of emergency plans and preparations in China, it is the first time to propose the deterministic and probability acceptance criteria for practical elimination in China at this stage. The radiological release acceptance criteria at this stage are suggested that there is no evacuate action beyond 3 km from the site boundary, and no sheltering action beyond 5 km. It is recommended that the DEC�B design requirement is less than 100 TBq equivalent 137Cs. This criterion is also the criterion for large radioactive release in the level 2 PSA. 3) Demonstration method of practical elimination was proposed in this paper, and demonstration of practical elimination for HPR1000 was evaluated. In the final analysis, DBA is effectively mitigated by defence�in�depth, and severe accident prevention and mitigation measures are considered sufficiently for HPR1000. Even if severe accident considered in design is happened, the containment could be intact. Practical elimination is realized on HPR1000.

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Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

HPR1000 is the Generation Ⅲ pressurized water reactor (PWR)technology independently developed by China. In order to deal with the steam generator tube rupture (SGTR) accident and prevent pressurizer overflow, HPR1000 adopts some new features such as the combination of active and passive safety system, the reduction of pressure setting value of safety injection system, rapid cooling at the secondary side of steam generator. In order to analyze the impact of new design characteristics of HPR1000 on sequence and thermalhydraulic phenomena in large break loss of coolant accident (LBLOCA), the numerical simulation of LBLOCA for HPR1000 was carried out using TRACE, which had been approved by United States Nuclear Regulatory Commission (NRC) as a best estimate system analysis code. The most challenging accident condition, namely combination of the most dangerous break location and the most dangerous size, were selected from the perspective of nuclear safety review. The LBLOCA sequence of HPR1000 was obtained and analyzed. The critical moments in the simulated LBLOCA process were compared with that of other typical commercial PWR such as CPR1000 and AP1000 in the sequence of accident and response strategies. According to the typical characteristics of thermalhydraulic phenomena, the accident process was then divided into four stages, namely blowdown phase, refilling phase, reflooding phase and long term cooling phase. The main thermalhydraulic phenomena in different accident stages except the long term cooling phase were identified and evaluated. The integral phenomena involved in the HPR1000 LBLOCA were depressurization of reactor coolant system (RCS), the coolant flow from core and intact RCS loop to broken loop, the safety injection flow and bypass flow of accumulator (ACC). While the local phenomena were mainly blowdown flow at break, countercurrent flow limitations (CCFL) in downcomer and other channels with complex geometry, heat transfer in the core, twophase flow and steam entrainment in the core, etc. The dominant factors during the accident process were pressure difference between the break and RCS, the pressure difference between ACC and RCS, core decay heat power and heat stored on the thick wall of RCS components. The results show that the main factors influencing the accident process are the mass flow rate of the break and the pressure setting value of accumulator in the LBLOCA of HPR1000. The accident sequence and phenomena are basically consistent with the existing commercial PWR nuclear power plants. The key phenomena identified based on the calculation results can provide technical support and reference for the phenomenon identification and ranking, scaling analysis, and nuclear safety review.

• A core reloading optimization scheme is developed. • Deep learning models were used as rapid evaluators of neutronic parameters. • Multi-Objective Genetic Algorithm used for the core reloading pattern optimization. • The reloading pattern for Cycle 2 of the HPR1000 reactor is optimized. • Cycle length has increased by 21 EFPD. The deep learning and multi-objective genetic algorithm were employed to optimize the fuel reloading pattern for HPR1000, a state-of-the-art nuclear power reactor designed and operated in China, also known as Hualong-1. In this study, the deep-learning algorithm was applied to establish the rapid evaluator for fuel-reloading patterns, for which the random samples between fuel-reloading patterns and corresponding key core parameters were generated by our home-developed nuclear-design code, named Bamboo-C. The advanced machine-learning platform TensorFlow was utilized for the deep-learning model. Then, the multi-objective genetic algorithm was applied to search the optimal fuel-reloading patterns, combined with the rapid evaluator to evaluate the key core parameters in a very short time. The DAKOTA toolkit was employed for optimization using a multi-objective genetic algorithm, for which the cycle length and power-peak factors were selected as the target parameters to establish the fitness function. For verification and application, the above method has been applied to the fuel-reloading optimization for Cycle 2 of HPR1000 operated in China. The optimization pattern results in an extension of the cycle length by about 21 EFPD (Effective Full Power Day), keeping all the key safety parameters satisfying corresponding safety criteria. [ABSTRACT FROM AUTHOR]

• The transient thermal hydraulic performance of the Secondary Side Passive Residual Heat Removal Heat Exchanger (SPRS HX) in the HPR1000 reactor is simulated using computational fluid dynamics (CFD). • The C-shape tube bundle is modeled using the porous media approach along with the distributed resistance method. • The two-phase flow phenomenon in the cooling water tank (CWT) is simulated by the drift flux model. • Three dimensional distributions of fluid temperature, velocity and void fraction in the CWT are obtained. • The SPRS HX heat transfer capacity from the primary side to the secondary fluid is determined. In advanced pressurized water reactor HPR1000, the secondary side passive residual heat removal heat exchanger (SPRS HX) is a crucial component. It removes the decay heat using boiling and convection heat transfer in the event of accidents. The robust functioning of the SPRS HX significantly influences overall reactor safety. To investigate the thermal hydraulic attributes of SPRS HX, computational fluid dynamics (CFD) is used in this study. The tube region is simulated using the porous media method (PMM). To model the two-phase flow in the tank, the drift flux model (DFM) is employed. The assessment of heat transfer rate from the primary to the secondary sides is conducted by empirical correlations. The governing equations are addressed through the utilization of FLUENT. Specifically, the momentum and energy equations are augmented with additional source terms that account for flow resistance and heat transfer in the tube region. These modifications are enacted through User Defined Functions (UDF). Three-dimensional profiles of the void fraction, fluid temperature and velocity are acquired, allowing for a comprehensive analysis of the heat transfer attributes of tube region. [ABSTRACT FROM AUTHOR]

• During the accident, the total hydrogen release was 390 kg, with a maximum release rate of 0.9 kg/s. • The diffusion path of hydrogen in the containment is "source compartment → top space of the containment → bottom compartment". • The absence of hydrogen risk in most containment compartments demonstrates the effectiveness of the CHC system. • Post-inerting can reduce hydrogen risk in containment at severe accident. Limited HPR1000 operational data under accident conditions emphasize the importance of analyzing its Containment Hydrogen Combination System (CHC) for improving nuclear safety. GASFLOW is employed to simulate a large break loss of coolant accident (LB-LOCA) in the steam generator's hot leg section, assuming the Safety Injection System's active injection fails. Results show that the movement of hydrogen within the containment follows the pathway 'source compartment → top space of the containment → bottom compartment.' Initially, a distinct stratification of hydrogen occurs throughout the containment. Subsequently, the stratification phenomenon begins to manifest in the source compartment and the lower areas of the containment. The CHC system effectively mitigates the risk of hydrogen, prevents flame acceleration and the likelihood of deflagration-to-detonation transitions in most compartments. However, the source compartment still poses a risk. Additionally, injecting inert gas into the source compartment is an effective method to control the hydrogen risk within the HPR1000. [ABSTRACT FROM AUTHOR]

• PRS steam condensation characteristics was experimentally studied. • Test data cover full range of the prototypical pressure condition. • Effects of pressure, steam quality, and mass flow rate were studied. • Test data further expand database of steam condensation in tube. The heat exchanger is an important component of the passive residual heat removal system (PRS) in the Hualong Pressurized Reactor (HPR1000). The heat released via steam condensation inside the heat transfer pipes is delivered to water stored outside of the containment. The heat transfer along the heat transfer pipes is systematically studied experimentally. Three horizontal test sections with different dimensions are designed and employed in the experiment. The test data are obtained under different pressure, steam quality, and mass flow rate conditions and compared with existing steam condensation correlation. The test results reveal that the heat transfer coefficient decreases from 41.5 kW/(m2·K) to 24.0 kW/(m2·K) with the pressure increased from 1.0 MPa to 8.3 MPa. In the higher pressure conditions, the larger condensed flow rate provides larger thermal resistance, decreasing the condensation heat transfer coefficient. The steam condensation heat transfer coefficient is increased from 14.3 kW/(m2·K) to 41.4 kW/(m2·K) as the steam quality is increased from 0.07 to 0.70. The condensation heat transfer coefficient is increased from 28.6 kW/(m2·K) to 45.8 kW/(m2·K) as the mass flow rate is increased from 0.029 kg/s to 0.072 kg/s. The maximum prediction deviation of Shah's correlation for all test data is ±34.9 %. [ABSTRACT FROM AUTHOR]

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The calculation model of HPR1000 has been built with MAAP code. The key parameters of fission products release, transfer and deposition have been screened out, and 500 input cases have been generated by Latin-Hypercube sample. The probability distributions of release fraction under the condition of intact containment for each fission product groups have been calculated. The result from uncertainty analysis has been used to calculate the off-site dose rate based on Zhangzhou NPP meteorology conditions with different percentile FP release. It has been demonstrated that no sheltering is needed beyond 3 km from the reactor even with conservative consequences.

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Copyright of Coatings & Protection / Tuceng yu Fanghu is the property of Coating & Protection Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 (Lerchl et al., 2016). By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR), and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN-CNPTRI (China General Nuclear—China Nuclear Power Technology Research Institute) with their own code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

• The validation of the HPR1000 using the SHARK code is presented in this paper. • The ARO case, R case, and SB case are simulated by SHARK, with errors of 148 pcm, 3.7%, and 4.7% in the ARO, R, and SB cases, respectively. SHARK is a new whole core transport code developed by the Nuclear Power Institute of China (NPIC). The constructive solid geometry (CSG) is used in SHARK to model the reactors. The subgroup method based on an equivalence cross-section interpolation table is applied to treat resonance self-shielding effects. The 2D/1D transport methods can be chosen in SHARK. The HPR1000 which is a 3rd commercial reactor has independent intellectual property owned by China. The validation of SHARK is processed via the HPR1000 in this work. The numerical results demonstrate that SHARK can calculate whole core directly and accurately. [ABSTRACT FROM AUTHOR]

The advanced Gen-III pressurized water reactor (PWR) HPR1000 utilizes rhodium self-powered neutron detectors (SPNDs) for reactor monitoring to guarantee safety operation. However, the neutronic characteristics of rhodium SPND result in its sensitivity being dependent on both thermal neutrons and epithermal neutrons and thereby introducing uncertainty. In this study, a computational model based on the Monte Carlo (MC) method is presented for the neutron sensitivity of rhodium SPND, and the calculated values are compared with experimental measurements to prove its effectiveness. Then several fuel assemblies at symmetrical locations in the HPR1000 are selected to quantifiably evaluate the sensitivity uncertainties of rhodium SPNDs. The calculation methods and evaluation results have successfully provided guidance for industrial applications in the HPR1000 and could be a valuable reference for other reactors. • [a]Rhodium SPND is designed to monitor reactors effectively due to strong signals. • [b]Neutronic features perform sensitivity uncertainty and are analyzed. • [c]An exact model is proposed to calculate neutron sensitivity with errors within 2%. • [d]Sensitivity variations could up to 8.0% under different in-core conditions. • [e]Calculated results successfully applied in nuclear energy industrial applications. [ABSTRACT FROM AUTHOR]

• Natural circulation characteristics of the PRS system was studied. • Sparger was designed and installed to alleviate the flow oscillation. • The effect of the sparger was tested on the integral test facility. • The sparger can effectively improve PRS safety performance. The Hualong pressurized water reactor (HPR1000) utilizes a secondary passive residual heat removal system (PRS) to maintain the core in a safe state within 72 h after a plant blackout incident (SBO). Two emergency make-up tanks (EMT) are incorporated into the PRS to provide safe injection to the secondary side of the steam generator (SG). However, during the early stages of PRS operation, direct contact condensation (DCC) of steam in the EMT can cause injection flow oscillations, thereby delaying the EMT safe injection process. The transient experiment was conducted on the integral test facility ESPRIT to study the DCC and injection flow oscillation characteristics of the EMT. Furthermore, the steam sparger was designed and installed on the inlet of the EMT to study its mitigation effect on the injection flow oscillation. The test results reveal that the DCC in the EMT brings about the water hammer and prevents gravity injection. The sparger transfers the momentum of the ingress steam from vertically to horizontally and the saturation layer formed at the steam-liquid interface is therefore well preserved. The injection flow oscillation is thus suppressed. After the sparger is installed, the average injection flow rate is higher than that in the original condition by 3.4 %, and the pressure is slightly lower than that of the original condition by averaging 0.08 MPa. The marginal improvement introduced by the sparger can facilitate system depressurization and improve PRS safety performance. [ABSTRACT FROM AUTHOR]

The mechanistic analysis of debris bed melting is particularly important to understand the severe accident scenario and to improve the development of integral severe accident analysis codes. In previous work, we had successfully predicted the transient melting process of debris bed in the hypothetical severe accident of HPR1000. However, there was a problem of high computation cost by using large eddy simulation (LES) model, which was not conducive to the fast calculation of analysis codes. Therefore, a simplified calculation model based on the phase change effective convectivity model was developed to improve calculation efficiency, and both the COPRA experiment and the numerical study using LES model were utilized to validate the reliability of the developed model. Results indicated that the predictions by the simplified calculation model agreed well with the data of COPRA experiment and the numerical study using LES model with the maximum deviation of 10.93%. Moreover, the calculation efficiency of the simplified calculation model was improved by at least 320 times compared to the previous LES model, suggesting that the simplified calculation model was effective to enhance the calculation efficiency and ensure the reasonable simulation accuracy. Additionally, sensitivity analysis of stainless-steel content in debris bed indicated that the increasing stainless-steel content resulted in the lower maximum temperature, smaller peak heat flux and thicker final crust thickness. Sensitivity analysis of zirconium oxidation fraction in debris bed indicated that the increasing zirconium oxidation fraction rendered the maximum temperature and peak heat flux of debris bed to rise, but there was no large change in the final crust thickness. The debris bed with 50% stainless-steel content and 80% zirconium oxidation fraction achieved the overheat with superheat of 96.7 K and the highest heat flux of 0.44 MW m−2. These findings provided a reference for the research of debris bed melting and the developments of integral severe accident analysis codes. [ABSTRACT FROM AUTHOR]

[Display omitted] • e Improvements on hydrogen risk assessment for NPP by combing lumped parameter and CFD codes • Uncertainty cases considering from both perspectives of time step and spatial dimensions • Providing a new direction to enhance the safety of NPP on the topic of hydrogen risk The risk of hydrogen combustion after the Fukushima nuclear accident has always been a topic of concern for the nuclear power industry. In the hydrogen risk assessment of probability safety analysis (PSA) in nuclear power plants, the traditional methods by using lumped parameter program method is fast but may have large uncertainties, while the newly developed CFD analysis method is more accurate but has not yet been widely applied to the hydrogen risk analysis of PSA due to a variety of factors. The lumped parameter analysis program MAAP is used firstly to obtain the hydrogen parameters under the severe accident of small LOCA in this paper, based on China's third-generation large-scale pressurized water reactor (PWR) nuclear power plant HPR1000, and then the probability of deflagaration to detonation (DDT) for hydrogen risk is analyzed with uncertainty. Secondly, the CFD software GASFLOW program is used to analyze the same accident sequence to obtain more accurate hydrogen distribution and other parameters, and at the same time to obtain the DDT probability value of hydrogen risk. The hydrogen distribution obtained by CFD calculation can be used to guide the uncertainty value of the lumped parameter program to produce more accurate DDT value of hydrogen risk. The analysis results show that there is a certain uncertainty in hydrogen risk assessment when using lumped parameter program method, which can be corrected by combining the gas distribution analysis results given by CFD software, to obtain a more accurate and reliable probability value, so as to provide a more accurate reference for PSA analysis and to improve the overall safety of the nuclear power plant. [ABSTRACT FROM AUTHOR]

In this paper, a systematic uncertainty quantification of dynamic mode decomposition (DMD) for xenon dynamic prediction is brought out based on HPR1000 reactor. The DMD method is a data-driven approach that decomposes complex systems into spatio-temporal structures and can be used to predict the power distribution in the process of xenon oscillation. To further investigate and improve the robustness of DMD with respect to observation noise, different error metrics are established. The dependence of the prediction error on observation noise level and hyper parameters including the window length and the singular value truncation threshold are investigated. Various numerical experiments based on a typical fuel cycle of the HPR1000 reactor demonstrate that DMD with optimal hyper parameters is robust with respect to observation noise, which confirms that DMD is feasible for real engineering applications. • A systematic uncertainty quantification of DMD is brought out for xenon dynamic prediction. • The dependence of the error on observation noise level and hyper parameters are investigated. • Experiments on HPR1000 reactor demonstrate DMD's feasibility for engineering applications. [ABSTRACT FROM AUTHOR]

HPR1000 is an advanced nuclear power plant (NPP) with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. On one hand, it is an evolutionary design based on proven technology of the existing pressurized water reactor NPP; on the other hand, it incorporates advanced design features including a 177-fuel-assembly core loaded with CF3 fuel assemblies, active and passive safety systems, comprehensive severe accident prevention and mitigation measures, enhanced protection against external events, and improved emergency response capability. Extensive verification experiments and tests have been performed for critical innovative improvements on passive systems, the reactor core, and the main equipment. The design of HPR1000 fulfills the international utility requirements for advanced light water reactors and the latest nuclear safety requirements, and addresses the safety issues relevant to the Fukushima accident. Along with its outstanding safety and economy, HPR1000 provides an excellent and practicable solution for both domestic and international nuclear power markets.

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