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The emergency passive residual heat removal system (EPRHR) in Hualong Pressurized Reactor (HPR1000) adopts natural circulation to remove the core residual heat passively during the event of feed line break or station blackout (SBO) accident. The Emergency Secondary Passive Residual heat removal system Integral Test facility (ESPRIT) was constructed to obtain the natural circulation characteristics of EPRHR under various initial and boundary conditions. Resistance matching was conducted in advance of the experiments, ensuring the same resistance characteristics between the test loop and the prototype. The natural circulation characteristics under different heat exchanger (HX) area, steam generator (SG) liquid level and system resistance were studied experimentally. The test data reveal that the system pressure/temperature rises dramatically with the decline of the HX heat transfer area when the HX heat transfer area is small. The natural circulation flow rate increases with increasing of the SG power, and with decreasing of the HX heat transfer area in the experiment. The natural circulation flow rate oscillates with certain frequency in low power rate conditions. The SG liquid level has no significant effect on system pressure/temperature and natural circulation rate in the experiment. The pressure decline rate decreases with the increase of the resistance coefficients. The condensed water flow rates decrease with the increasing of the resistance. [ABSTRACT FROM AUTHOR]

Steam Generator Tube Rupture (SGTR) accident was reclassified as Class III Design Basis Accident (DBA) in 1991 due to its higher frequency of occurrence. Under Class III condition, radioactive release safety criteria are even stricter than thermal hydraulic criteria. In this paper, HPR1000 SGTR accident was simulated with two kinds of mitigation measures after steady-state initialization. Key system responses (such as pressure, mass flow rate in primary and secondary system, SG level, short term phase terminating time and radioactive release) were analyzed and compared for the two strategies. No SG overfilled nor other thermal hydraulic limit occurred for both strategies, while lower SG overfill risk and higher radioactive release risk existed for strategy 1 and converse situation existed for strategy 2. Specially, it is recommended VDA to be actuated by SG level, and instant main feedwater and auxiliary feedwater isolations were strongly required for strategy 2. When designing SGTR mitigation measure, there should be a compromised balance between SG overfill risk and radioactive release. Optimized mitigation strategy can be obtained through more accurate activity signal and higher auxiliary feedwater mass flow rate. [ABSTRACT FROM AUTHOR]

All advanced NPPs are required to improve safety strategy to cope with design extension condition (DEC) after Fukushima accident. One design improvements for HPR1000 is adoption of secondary Passive Residual Heat Removal System (PRS) to cope with station blackout accident together with loss of all feedwater, which can help to strengthen defense in depth capability and enrich DEC mitigation measures. To make the best use of PRS decay heat removal capability, investigation of PRS startup performance and decay heat removal capability after MFLB accident are investigated based on HPR1000 RELAP5 model. PRS startup response featured by pressure, mass flow rate and void fraction are investigated under three startup strategies. Strategy 3(backflow valve open strategy) is recommended due to minimum fluctuation during initiating stage, in which strategy steam isolation valve is normally open and backflow isolation valve is used to initiate PRS. Then Main Feedwater Line Break (MFLB) accident was simulated with the assumption of loss of auxiliary feedwater. Response characteristics in primary loop and PRS are investigated. After PRS being initiated after scram with 60 s delay, natural circulation in PRS will establish rapidly, and decay heat will be removed under natural circulation operation mode. No overall boiling occurred in primary circuit before isolation of affected SG. Secondary PRS is adequate to remove decay heat after DEC. [ABSTRACT FROM AUTHOR]

The steam generator secondary side emergency passive residual heat removal system (EPRHR) is innovatively introduced into the Hualong Pressurized Reactor (HPR1000) to enhance its safety performance in station blackout (SBO) scenarios. The Emergency Secondary Passive Residual heat removal system Integral Test facility (ESPRIT) was constructed to obtain the safety performance of EPRHR under various initial and boundary conditions. The scaling ratio of power/volume between ESPRIT and HPR1000 is 1:62.5. The resistance matching experiments were conducted in the first place to restore the prototypical pressure drop characteristics of the EPRHR. Three transient experiments, each with the duration of 72 h, were conducted under prototypical, elevated heating power and elevated resistance conditions to evaluate the safety performance of HPR1000 under SBO conditions. The experimental results reveal that the steam generator pressure decreases dramatically in the early stage after the startup of EPRHR. The safety injection rate decreases dramatically in the elevated resistance condition compared with the prototypical one. The liquid level in the cooling water tank decreases by 5.4% in the steam generator power elevation condition. The designed volume of cooling water tank satisfies the heat removal demand in SBO accident within 72 h. The heat sink possesses sufficient cooling ability under elevated power condition. The system temperature remains decreasing with the decrease of residual heat in each of the transient experiment. It is also found that effective natural circulation and continuous heat exchanger cooling always remain in three experiments. [ABSTRACT FROM AUTHOR]

To cope with SBLOCA in absence of High-Head Safety Injection (HHSI) from design of HPR1000, atmospheric relief system (originally named as VDA in French) is uniquely designed to help to trigger Middle Head Safety Injection (MHSI) or Low Head Safety Injection (LHSI) earlier through cooling primary system quickly after SBLOCA. To make the best use of VDA decay heat removal capability, primary and secondary system of HPR1000 was modeled with RELAP5/SCDAP computer code. After steady-state initialization, a cold leg 30 mm break SBLOCA was simulated with six simulation conditions and five additional cases including availability of ACCU, different VDA discharge locations and area. Response characteristics of primary loop under different VDA working conditions are investigated. Pressurizer pressure decreases rapidly to lower level to trigger the reactor scram, VDA activation and accumulator safety injection sequently. Peak cladding temperature is 899.45 K occurring at 222 s, which is far below the safety limit. Activation of VDA can trigger ACCU SI earlier with a critical point, while positive reactivity will be introduced due to negative moderator temperature effect and Doppler effect. Larger VDA discharge capability will introduce larger reactivity feedback, as well as induce lower core level and SG level. It's suggested that VDA discharge condition should be chosen before the critical point, with the compromise with reactivity feedback introduced due to the negative moderator temperature effect. [ABSTRACT FROM AUTHOR]

• A neutronics method for fuel-assembly bowing modeling has been proposed. • Our home-developed code Bamboo-C based on "two-step" scheme has been improved. • Fuel-assembly bowing cases of HPR1000 have been designed and verified. The fuel-assembly bowing is widely observed in PWR, due to the complex operational environment in the core. After the fuel-assembly bowing, the three-dimensional (3D) water-gap distributions around the fuel assemblies would be deviated from the nominal values. From the view of neutron physics, the variations of water gap would result in over-moderation or under-moderation of the neutrons locally, hence impact the results of the neutronics simulations. In order to estimate the neutronics effects of the fuel-assembly bowing, our home-developed Bamboo-C code system based on the two-step scheme has been improved. In the lattice code LOCUST, the few-group constants were generated with different sizes of water gaps around the fuel assembly. In the core code SPARK, the neutron-diffusion solver has been improved to consider the geometry-grid variations due to the whole-core fuel-assembly bowing. For the method verifications, two test cases of HPR1000 with different kinds of fuel-assembly bowings have been designed and simulated, with corresponding reference results provided by the Monte-Carlo code NECP-MCX. The verification results show that well-agreed eigenvalue and assembly-averaged power distributions can be provided by the improved Bamboo-C code. [ABSTRACT FROM AUTHOR]

• A long-term passive cooling system is proposed for HPR1000. • The heating power is below 9 MW, while the T ave of the pool water is below 90 °C. • The position of the tube bundle affects the T ave , and when γ = 3/4, T ave is the smallest. The EPRHR in the HPR1000 adopts natural circulation to meet the heat removal requirements of the containment for 72-hours under accident conditions. To reduce the evaporation depletion of water in CWT and enhances the passive cooling capability of the EPRHR, a long-term passive cooling scheme based on a heat pipe heat exchanger is proposed for CWT in this paper. Experimental study and numerical simulation are used to study the natural convection characteristics in the CWT. The results show that the scheme can take away 9 MW heat load as the average water temperature (T ave) is below 90 °C. The heat removal ability of the tube bundle is weakened row by row and the maximum difference of heat removal ability between tube bundles is 23%. The position of the tube bundles affects the T ave , and when the dimensionless layout factor (γ) = 3/4, T ave (346.34 K) is the smallest. [ABSTRACT FROM AUTHOR]

Under normal operation condition of nuclear power plants, there may exist some special phenomena related to high burnup fuel cladding, such as breakaway oxidation, inner surface oxygen ingress and hydrogen pickup embrittlement. The effects of these phenomena on cladding fracture limit during LOCA conditions were investigated by fuel cladding research program sponsored by U.S. NRC. Based on the research, new LOCA acceptance criterion 10CFR50.46c draft was agreed and released. In the present study, research on compliance between large break LOCA of HPR1000 and new LOCA acceptance criterion 10CFR50.46c was performed. Firstly, a hydrogen content prediction model for different cladding material was innovatively proposed and then large break LOCA analysis of HPR1000 at different burnup points was performed. Research showed that the analysis results of large break LOCA of HPR1000 could meet requirement of 10CFR50.46c. The PCT and transient ECR margin were 190.1 °C and 10.1% respectively which were large enough for large break LOCA to ensure reactor core safety. • A hydrogen content prediction model for different cladding material was innovatively proposed. • LBLOCA analysis of HRP1000 was performed at different burnup points by adopting self-developed codes LOCUST and CATLAPA. • Research on compliance between LBLOCA of HPR1000 and 10CFR50.46c was performed for the first time. [ABSTRACT FROM AUTHOR]

• The nonlinear and transfer function models are developed for HPR1000 using the first-principles approach. • The models consist of the reactor coolant system, pressurizer, UTSG, and steam bypass system. • Four cases of typical transient simulations are performed. • Responses of transfer functions are compared with those obtained by the nonlinear model for the four cases. The purpose of this paper is to demonstrate a transfer function development process applicable to pressurized water reactors, specifically for the HPR1000. Firstly, the modeling methods using conservation equations are adopted to obtain the nonlinear dynamic model of the reactor coolant system, the pressurizer, the U-tube steam generator, and the steam bypass system. Then, the linearization of each dynamic model is performed based on the small perturbation theory. Laplace transformation is carried out to the linearized dynamic model to the transfer functions. Finally, transient simulations with the nonlinear models are executed by introducing typical disturbances to verify the transfer functions. The simulation results show good agreement and prove the proposed transfer function development method. Therefore, the developed transfer function can predict the dynamic behaviors of HPR1000 and be used for control studies. The methodology is a reference to other similar reactors for transfer function development. [ABSTRACT FROM AUTHOR]

Residual heat removal system (RHRS) is one of the important and complicated safety systems for Hualong Pressurized Reactor 1000 (HPR1000). Therefore, it is essential to carry out the reliability assessment of the RHRS for HPR1000. In order to fill the gap of how to assess the reliability of the RHRS with uncertainty, the failure mode and effect analysis (FMEA), fault tree (FT) analysis (FTA), and fuzzy Bayesian network (BN) (FBN) methods were applied to establish a novel approach for assessing its reliability. The proposed approach included: (1) identifying the events containing failure modes, failure causes, and resulted failures for the RHRS by FMEA; (2) analyzing the deterministic logical relationship among events, and establishing the FT for the RHRS by FTA; and (3) establishing the FBN model for reliability assessment of the RHRS by FBN according to the transformation process from FT to FBN model, and modifying the conditional probability tables (CPTs) so as to express the event polymorphism and non-deterministic logical relationship between events. The applicability of the approach was validated in East China. • A novel approach for reliability assessment of the RHRS with uncertainty was proposed. • The failure probability of the RHRS was calculated. • The sensitivity analysis of the RHRS was carried out. • The uncertainty analysis of safety barriers was conducted. • The applicability of the approach was validated by a typical case. [ABSTRACT FROM AUTHOR]

• Experiment was conducted to obtain the steam direct contact condensation characteristics in makeup tank. • Three types of spargers were designed and their performance were evaluated. • The heat transfer coefficients were obtained and the effects of spargers were studied. • The steam flow rate and pressure oscillations were analyzed in the experiment. • The spargers could enhance the safety performance of passive safety system in accident scenarios. The emergency makeup tank (EMT) is an important component of passive residual heat removal system (PRS) in Hualong Pressurized Reactor (HPR1000). However, the steam direct contact condensation (DCC) occurred in the EMT brings pressure and injection flow oscillation in station blackout scenarios, affecting the safety injection performance of EMT. Experimental facility was constructed to obtain the steam DCC characteristics under various initial and boundary conditions. The experimental phenomenon was observed and test data were obtained and analyzed. Three types of spargers were designed and installed to alleviate the pressure and flow rate oscillation in EMT. The test results reveal that spargers can significantly reduce the DCC duration and promote the water thermal stratification in EMT. The heat transfer coefficient of steam condensation substantially decreases after the spargers are installed. The steam flow rate and pressure oscillates vigorously at the early stage of the sparger-free experimental condition, and the oscillations are greatly reduced after the spargers are installed. In EMT drain experiment, the maximum steam flow rate and minimum pressure are reached simultaneously when the steam supply rate is balanced with the steam condensation rate. The drain of water begins at the start of the experiment while the liquid level in the EMT is not reduced due to the water supplement via rapid steam condensation. The EMT evacuation duration is significantly reduced when the spargers are installed, which could enhance the safety performance of passive safety system in accidental scenarios. [ABSTRACT FROM AUTHOR]

• The dynamic mode decomposition (DMD) is adapted to forecast the power distribution during the process of xenon oscillations. • The DMD based strategy is a fast, model-free and easy-to-implement data-driven method. • Numerical results on HPR1000 shows the efficiency in the prediction of power distribution. • The current amount of SPNDs in HPR1000 is enough to capture the dynamical behavior of xenon oscillations. Detection and prediction of xenon induced oscillations are an important part in the operation of pressurized water reactors. Several models have been proposed for the prediction or estimation of xenon oscillations with drawbacks e.g. strongly depend on the initial xenon and iodine distributions, hard-to-implement or computationally costly. In this article, we proposed a fast, model-free and easy-to-implement data-driven strategy based on dynamic mode decomposition (DMD) to forecast the power distribution during the process of xenon oscillations. Various comparative experiments based on HPR1000 reactor show that the proposed data driven strategy is able to capture the complex relational characteristics of the temporal and spatial data of xenon oscillations. It can efficiently discover the hidden dynamicities and thus offers an accurate prediction of the system behavior. [ABSTRACT FROM AUTHOR]

• Single failure criteria and operator mitigation intervention after HPR1000 SGTR were investigated. • VDA stuck open and AFW isolation failure was conservative single failure criteria for two strategies respectively. • Strategy 1 can provide multiple redundant engineered safety features from T-H safety aspect. Establishment of a unified single failure criteria hypothesis and operator action for Steam Generator Tube Rupture (SGTR) accident in Deterministic Safety Analysis (DSA) is not an easy thing due to the complexity of the transient and the various operator actuations. Also, it's very difficult to finish leakage in 30 min without operator action. In HPR1000 type nuclear power plant design, two kinds of mitigation strategy to cope with SG overfill risk and radioactive release risk can be adopted. In this paper, behavior of HPR1000 under different operator actions and single failure criteria hypothesis for this two mitigation measures were investigated. System response featured by primary pressure, SG level, break mass flow rate, safety injection mass flow rate was obtained to investigate single failure criteria and operator manual intervention effect. For strategy 1, if VDA stuck open was chosen as single failure criteria, primary loop would be over cooled and accumulator safety injection would increase with more drastically fluctuation. In strategy 2, AFW isolation failure was chosen as single failure criteria from SG overfill risk aspect. Affected SG would overfill if AFW isolation failure lasting over 5 min. Combined manual PRZR spray and SIS termination can obtain faster depressurization and leakage termination rate within controllable core boiling and acceptable radioactive release in strategy 1. However, in Strategy 2, manual PRZR spray depressurization and manual SIS termination have little effect on accident mitigation due to VDA comparatively large cooling ability and early intervention. From T-H safety aspect after SGTR, operator manual intervention has more obvious effect in strategy 1 than in strategy 2, which can provide multiple redundant engineered safety features. [ABSTRACT FROM AUTHOR]