Three-loop Pressurized Water Reactor Research Articles

The scaling effect analysis of counterpart test between LSTF and ATLAS IET facilities on 1% top head break LOCA

ATLAS B5.1 test is a 1% small break loss of coolant accident (SBLOCA) at the top of the reactor pressure vessel with a total failure of a high-pressure injection (HPI) system. It was conducted as a counterpart test of LSTF SB-PV-07. The LSTF is scaled from a three-loop pressurized water reactor (PWR) Westinghouse design, while the ATLAS facility is scaled from APR1400. Thus, to determine B5.1 test conditions, a scaling analysis has been conducted between the two test facilities of LSTF and ATLAS. We reviewed the scaling method used for the ATLAS facility and found the initial and boundary condition for the counterpart test. Then, we compared the experimental data of the ATLAS B5.1 test against that of LSTF SB-PV-07. To reduce the scaling distortion between two IET facilities, we modified the RELAP5 input of the ATLAS facility toward the LSTF facility considering the elevation difference and design difference. The internal design difference between the ALTAS and LSTF facilities generated a significant difference in the plant transient beyond the scaling distortion.

Improvement of the uncertainty method with model treatment and optimized global sensitivity analysis and its application to pressurized water reactor

Best estimate plus uncertainty methodology is an advanced technology for licensing application of the nuclear power plants, but some aspects can still be improved in the conventional input-driven methodologies. For instance, uncertainties of the closure models are not properly handled; local sensitivity analysis methods not suitable for complex nuclear systems are still widely utilized; the sensitivity analysis process needs to be performed again every time new parameters are identified. Optimization researches were carried out in the study. A structural method was proposed to quantify the uncertainties of the closure models, and a low-cost moment-independent global sensitivity analysis method was developed to improve the sensitivity analysis framework. By integrating the improvements into the conventional input-driven framework, an uncertainty method with model treatment and improved sensitivity analysis was obtained. Finally, the method was validated through a large break loss of coolant accident in a three-loop pressurized water reactor.

Optimization of sensitivity analysis in best estimate plus uncertainty and the application to large break LOCA of a three-loop pressurized water reactor

The best estimate plus uncertainty (BEPU) methodology is developed to carry out realistic safety analysis of the nuclear power plant. However, the sensitivity analysis in the conventional methodologies often leads to inaccurate results when lacking of the phenomenon identification and ranking table (PIRT) or adopting the local sensitivity analysis methods. As the initial inputs determined in short of the PIRT may neglect some important parameters or identify some unimportant parameters, and the local methods are not appropriate for complex and non-linear systems. In this study, an optimized moment-independent global sensitivity analysis method with very low computational cost was utilized to improve the sensitivity analysis framework of the BEPU methodology. The reliability as well as the applicability of the method were assessed through some tests and a large break loss of coolant accident (LBLOCA) of the LOFT facility. The improved sensitivity analysis framework was then applied to the BEPU analysis of a typical three-loop pressurized water reactor CPR1000, and the reliability of the optimized moment-independent method was further confirmed by comparing the sensitivity analysis result of the method to the results of the BEMUSE project and the Spearman method. Besides, two new parameters were identified to have great influence on the PCT, and the two parameters should be taken into consideration in the uncertainty analysis of the BEPU methodology.

Risk Reduction Assessment for Installing RCP Shutdown Seal

As one of the lessons learned from the Fukushima Daiichi accident, long-term station blackout (SBO) and subsequent loss of ultimate heat sinks have prompted discussion on this topic in the nuclear industry. The SBO sequences for a Westinghouse three-loop pressurized water reactor (PWR) have been under investigation for a long period. To cope with the long-term SBO issue, many nuclear power plants have decided to replace the reactor coolant pump (RCP) seal by the new passive thermal shutdown seal (PSDS). The PSDS is a fail-safe protection device that will significantly reduce leakage from the RCP seal in case of loss of cooling. This makes a seal loss-of-coolant accident no longer a risk-significant event, and the relevant probabilistic risk assessment (PRA) models need to be modified to reflect the associated plant change. The PRA model of a Westinghouse PWR plant has been reviewed to reflect this more strongly; i.e. loss of component cooling water (CCW), loss of 4-kV vital alternating-current power, and loss of off-site power are revised for their sequences. According to the Westinghouse analysis, the PSDS temperature must be maintained below 104°C, and the operators have to control the RCP speed. In this paper, those factors are incorporated into the event tree structure revision of the loss of CCW event (TC) and loss of power either off-site (TP) or vital power A train (TAPB). In another case, LOOP initiating events need to consider the time span that the blackout conditions would affect the RCP seal integrity. Because the RCP will be tripped automatically, the limitation associated with RCP speed will not be applicable. Compared with the TC, TP, and TAPB event tree base cases, RCP speed slowdown or available time span is introduced into the PSDS model. The relevant part in the PRA model is subject to review and modification. The risk reduction associated with the PSDS is found to be significant.

Risk-Informed Safety Analysis for Accident Tolerant Fuels

Accident tolerant fuels (ATFs) are being tested by different nuclear vendors and research organizations, and their introduction into the U.S light water reactor fleet is planned for the second half of the 2020s. Under the framework of the U.S. Department of Energy Light Water Reactor Sustainability (LWRS) Program, as part of the LWRS Risk-Informed Systems Analysis Pathway, research is being conducted at the Idaho National Laboratory (INL) to develop tools and methods that can help the industry to quantify the benefits from adopting ATF technology. In this paper we describe the developed risk-informed methodology including the safety analysis code improvements, and we present some results for selected accident scenarios. The developed methodology combines the INL state-of-the-art deterministic Best Estimate code RELAP5-3D and the probabilistic risk analysis tools RAVEN and SAPHIRE. The analyses are performed on a three-loop pressurized water reactor, simulating station blackout and large-break loss-of-coolant accidents and considering near-term ATFs or iron-chromium-aluminum and chromium-coated clads. Finally, we show how, applying our methodology, the new core damage frequency (CDF) can be assessed. The results indicate that the main benefit in introducing near-term ATFs is a significant reduction in hydrogen production during accident conditions. No significant CDF reduction was found.

Preliminary optimization of emergency core cooling system with accident-tolerant fuels

Based on the excellent performance of accident-tolerant fuels (ATFs), the optimization schemes for the safety system of Chinese advanced three-loop pressurized water reactor during loss of coolant accident (LOCA) were proposed and corresponding feasibility analysis was carried out in this study. The modified system analysis code was used to safety evaluation and four ATF systems were investigated. Several modifications were made to the rod geometries in partial ATF systems to make up the poor neutron economic performance of FeCrAl and MoMAX. The emergency core cooling system (ECCS) was optimized during LOCA with different break sizes. The results show that it is feasible for all ATF systems to cancel the high pressure safety injection (HPSI) or cancel both HPSI and low pressure safety injection (LPSI) but add an extra accumulator (ACC). Also the safety of reactors fueled with ATFs can be guaranteed with only 0.35 times the original flowrate of LPSI during large break LOCA (LB-LOCA). In the scheme of replacing the original ECCS with two extra ACCs during LB-LOCA, reactors loaded with partial ATFs are safe, while UO2-Zr systems are in danger as a contrast. This work could provide reference for the design of reactor safety system in the future.

Preliminary neutronic/thermal-hydraulic evaluation and safety system optimization of PWR loaded with fully ceramic microencapsulated fuel

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future.

TRACE simulations of IIST cooldown experiments

Abstract IIST (Institute of Nuclear Energy Research Integral System Test) facility is a RHRP (Reduced-Height and Reduced-Pressure) facility for the three-loop PWR (Pressurized Water Reactor) NPP (Nuclear Power Plant) Maanshan. This test facility was established in 1992 for safety studies of the Maanshan NPP. TRACE is an advanced thermal hydraulic code which is still developed and updated by U.S. NRC. So, validation of specific models against experiments is still necessary. In this paper, cooldown experiments in IIST facility are used to validate the TRACE models. Additionally, the results of TRACE calculations are compared to results of RELAP5 calculations. This comparison shows that the calculations of TRACE are in good agreement to the data of IIST experiments and to the calculational results of RELAP5. So, TRACE can be used to simulate cooldown transients.

Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

The American Society of Mechanical Engineers (ASME) Section III Rules for Construction of Nuclear Facility Components subsection NB-2331 Material for Vessels requires that the effects of irradiation shall be considered on material toughness properties in the core belt line region of the reactor vessel. The code also states that “the design specifications shall include additional requirements, as necessary, to ensure adequate fracture toughness for the service lifetime of the vessel.” In a design report of nuclear pressure vessel, the design and service loads do not include loads that are affected by fracture toughness of the material. However, in the cases of fitness-for-service assessment for component flaws (prevalent with age of component), irradiated material properties become highly relevant. An example of a fitness-for-service is that of a beyond design basis reactor vessel head drop accident in a pressurized water reactor with a nozzle junction flaw. As a case study, the critical size of a postulated external surface semielliptical circumferential crack in the combustion engineering three-loop pressurized water reactor nozzle–vessel junction is calculated using ansys Workbench (Academic version) with the applied impact load from the vessel head drop accident. Failure assessment diagrams for numerous crack depths and lengths were developed considering the fracture toughness properties of the irradiated reactor vessel steel. The mode I stress intensity results used in the failure assessment diagram were compared with the available finite element and the American Petroleum Institute (API) standard API 579 analytical solutions for validation, showing good agreement. From this case study, it is demonstrated that the effects of irradiation on fracture toughness become prominent at the same postulated crack size in the nozzle–vessel junction dispositioned as “safe” becomes “unsafe” in the fracture assessment diagram. Using the unified curve method, the irradiated fracture toughness data in design specification can be supplied so that it may be used in fitness-for-service analysis to account for component aging.

Risk Management for Seal Leakage Modeling of Station Blackout Core Damage Frequency

Station blackout (SBO) sequences for a Westinghouse-designed three-loop pressurized water reactor (PWR) with large dry containment are investigated. Recovery of alternating-current power is considered under two separate headings in event trees of loss of off-site power: recovery of the off-site power and recovery of the emergency diesel generators (EDGs) or diesel generator 5. A reactor coolant pump seal leakage model under SBO conditions for a Westinghouse PWR with high-temperature O-rings is used. Seal failure mechanisms, including popping open, binding, and O-ring extrusion, are considered. Success criteria are established based on the results of thermal-hydraulic calculations via the MELCOR and/or RELAP-5 codes. Furthermore, the WinNUPRA software package is used for sequence quantification.Three primary models are addressed: Basic, WOG2000, and Ultimate Response Guideline (URG). The core damage frequency (CDF) of SBO sequences for the Basic model is found to be very optimistic and underestimated. The improvement on CDF by introducing the URG model is found to be dependent primarily on the reliability of plant operators in performing the procedure URG under SBO situations.Two sensitivity analyses on the passive shutdown seal (PSDS) are performed: WOG2000 (PSDS) and URG (PSDS). The low-failure-probability PSDS results in a >94% improvement in CDF due to SBO since the CDF contribution from long-term SBO sequences becomes negligible. Furthermore, the URG strategies are shown to have a significant impact on the reduction of CDF since seal leakage is no longer a concern in the presence of the PSDS.The way that power recovered is distinguished from off-site power or on-site EDGs and the way that SBO CDF is assessed make the present study more realistic than general SBO models.A key analysis of the confidence bands of the SBO CDF with the aid of the uncertainty measure is also performed to observe the induced effects by the probabilistic seal failure modes and PSDS.

Uncertainty Analysis of the Conditional Exceedance Probability Calculation for a Probabilistically Significant SBO Sequence

The Risk Informed Safety Margin Characterization methodology combines traditional probabilistic safety assessment (PSA) and the best-estimate plus uncertainty approach. Consequently, both stochastic uncertainty and epistemic uncertainty can be taken into overall consideration to evaluate the risk-informed safety margin. Generally, in calculation of the event sequence success criteria in traditional PSA, the result can only be either success (zero) or failure (unity), which is because uncertainties are not properly taken into consideration. In this paper, the conditional exceedance probability (CEP) of a probabilistically significant station blackout sequence of a typical three-loop pressurized water reactor was calculated with the consideration of both stochastic and epistemic uncertainties by using RELAP5. To get the probability density function of the peak cladding temperature (PCT) of a particular sequence and corresponding CEP, random sampling analysis of major plant status parameters and stochastic parameters was performed. It is assumed that the core is damaged when the PCT reaches 1477.6 K. Through the calculation of CEP of this specific sequence, it can be found that core damage will take place in a certain possibility between zero and unity when taking plant status uncertainties and stochastic uncertainties into consideration. Therefore, the core damage frequency (CDF) of any probabilistically significant sequence can be recalculated to get a more precise CEP. With the application of the computational risk assessment method, not only can the conditional CDF be reasonably reduced, but also the revised model can be made sensitive to a system design change of limited scope. Compared to the traditional PSA evaluation without uncertainty analysis, the CDF of the loss–of–heat sink dominant group can be reduced by a factor of 8.75 ( / ).

Autonomous operation algorithm for safety systems of nuclear power plants by using long-short term memory and function-based hierarchical framework

With the improvement of computer performance and the emergence of cutting-edge artificial intelligence (AI) algorithms, an autonomous operation based on AI is being applied to many industries. An autonomous algorithm is a higher-level concept than conventional automatic operation in nuclear power plants (NPPs). In order to achieve autonomous operation, the autonomous algorithm needs to include superior functions to monitor, control and diagnose automated subsystems. This study suggests an autonomous operation algorithm for NPP safety systems using a function-based hierarchical framework (FHF) and a long short-term memory (LSTM). The FHF hierarchically models the safety goals, functions, systems, and components in the NPP. Then, the hierarchical structure is transformed into an LSTM network that is an evolutionary version of a recurrent neural network. This approach is applied to a reference NPP, a Westinghouse 930 MWe, three-loop pressurized water reactor. This LSTM network has been trained and validated using a compact nuclear simulator.

Study on the reactor core barrel instantaneous characteristics in case of Loss of Coolant Accident (LOCA) scenarios for loop-type PWR

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios.

Verification of the Parallel Pin-Wise Core Simulator pCTF/PARCSv3.2 in Operational Control Rod Drop Transient Scenarios

Thanks to advances in computer technology, it is feasible to obtain detailed reactor core descriptions for safety analysis of the light water reactor (LWR), in order to represent realistically the fuel elements design, as is the case for three-dimensional coupled simulations for local neutron kinetics and thermal hydraulics. This scenario requires an efficient thermal-hydraulic code that can produce a response in a reasonable time for large-scale, detailed models. In two-fluid codes, such as the thermal-hydraulic subchannel code COBRA-TF, the time restriction is even more important, since the set of equations to be solved is more complex. We have developed a message passing interface parallel version of COBRA-TF, called pCTF. The parallel code is based on a cell-oriented domain decomposition approach, and performs well in models that consist of many cells. The Jacobian matrix is computed in parallel, with each processor in charge of calculating the coefficients related to a subset of the cells. Furthermore, the resulting system of linear equations is also solved in parallel, by exploiting solvers and preconditioners from PETSc. The goal of this study is to demonstrate the capability of the recently developed pCTF/PARCS coupled code to simulate large cores with a pin-by-pin level of detail in an acceptable computational time, using for this purpose two control rod drop operational transients that took place in the core of a three-loop pressurized water reactor. As a result, the main safety parameters of the core hot channel have been calculated by the coupled code in a pin level of detail, obtaining best estimate results for this transient.

Investigation of severe accident scenario of PWR response to LOCA along with SBO

Severe accident analysis is important for the safety evaluation of a nuclear power plant (NPP). In this paper, analysis has been performed for a Chinese three-loop Pressurized Water Reactor (PWR) severe accident induced by loss of coolant accident (LOCA) along with Station- Block-Out (SBO) using the MIDAC code. The simulation results show the influence of different break sizes in the cold leg on the severe accident progression.Important parameters, such as primary coolant system (RCS) pressure, core melting time, pressure vessel rupture time. are analyzed with three different break sizes in the cold leg. With the break size of 0.002 m2, the reactor core starts to melt at 1481 s and the lower head fails at 10,317 s after the accident occurs. With a larger break size of 0.05 m2, the reactor core starts to melt at 1066 s and the lower head fails at 2473 s. If the break size is further enlarged to 0.2 m2, the reactor core starts to melt at 422 s and the lower head fails at 1757 s. Another case in which a hot leg break is with the size of 0.002 m2 is performed. The reactor core starts to melt at 3641 s and the lower head fails at 14,744 s. It is less severe than the cold leg break accident.The severe accident prevention and mitigation measures are summarized based on the computation results. The results are helpful to develop the management strategies and guideline for the severe accident of the similar types of PWRs.

Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50kPag (46.92psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92kPag (45.24psig). But it is still below the design value of 413.69kPag (60psig). The calculated peak vapor temperature inside the containment is 187.0°C (368.59F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42°C (381.95F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. Since the results are similar with the Mananshan FSAR, and the applicability of the GOTHIC dry containment model is assessed.

Evaluation of the Impact of SAMG on the Level-2 PSA Results of a Pressurized Water Reactor

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will lower the containment failure frequency and reduce the amount of radionuclides released to the environment during the accident. The plant studied is the Maanshan NPP of Taiwan Power Company, which employs a Westinghouse-designed three-loop pressurized water reactor (PWR) with large dry containment.The containment system event trees and containment phenomenological event trees of the Level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The HCR (Human Cognitive Reliability) and THERP (Technique for Human Error Rate Prediction) models are used to quantify the human error probability (HEP) of all the actions in the Level-2 PSA model. The MAAP4 (Module Accident Analysis Program version 4) code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the frequency of most of the source term categories is reduced except the one in which both the reactor pressure vessel and containment are intact. The containment failure frequency is reduced by 14.8% after the implementation of SAMG. The frequency of containment early failure is reduced by 16.2%. Most of the reduction in the containment early failure frequency comes from the reduction in the induced steam generator tube rupture (STGR). The frequency of induced SGTR was reduced from 2.3 × 10–7/reactor·yr to 1.0 × 10–8/reactor·yr.

Effect of Highly Enriched/Highly Burnt UO2 Fuels on Fuel Cycle Costs, Radiotoxicity, and Nuclear Design Parameters

A study of high-burnup pressurized water reactor (PWR) fuel management schemes extending to 100 GWd/tonne is presented. The Studsvik Scandpower code suite was used to model a Westinghouse three-loop PWR core, and realistic loading patterns were derived. The loading patterns were optimized for minimum power peaking and maximum cycle length using engineering judgment and automated binary shuffles. Gadolinia was found to control power peaking to within current FΔH design limits up to 70 GWd/tonne, with only a slight deterioration thereafter. The moderator temperature coefficient, boron coefficient, and control rod worth were calculated and shown to fall within existing design limits.An economic analysis was carried out to determine the most economic discharge burnup based on fuel cycle costs only. It was found that the lowest fuel cycle costs were obtained with average discharge burnups between 70 to 75 GWd/tonne (initial enrichments between 6 to 7 wt%).The energy generated per tonne of uranium ore used was calculated and shown to peak between 40 to 60 GWd/tonne. Also, the radiotoxicity generated per GW·yr(electric) was calculated for each fuel management scheme and found to reduce considerably with burnup between 100 and 100 000 yr.

Safety Evaluation of Primary-Side Bleed-and-Feed Scenarios by an Experimental Loss of Feedwater Event at the IIST Facility

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on a total loss-of-feedwater (LOFW) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test (IIST) facility. Four experiments were conducted with a different number of pressurizer power-operated relief valves (PORVs) and centrifugal charging pumps (CCPs) engaged in performing the primary-side bleed-and-feed operation. In addition, the effect of opening a pressure vessel venting-path was also evaluated. The focus was on the effectiveness and adequacy of current EOPs and the observation of associated thermal-hydraulic phenomena. The test results indicated that the current EOPs can function well in mitigating the consequences of a LOFW event provided that at least two fully opened PORVs and one actuated CCP are used. The opening of a pressure vessel venting path during the bleed-and-feed operation is helpful in the primary system depressurization but has little effect to increase the pressure vessel coolant inventory. However, by opening one PORV, precaution for the reactor coolant system (RCS) integrity should be taken for system overpressurization caused by insufficient heat removal capability. Furthermore, an effective control of restored feedwater to a low injection rate is favorable to minimize the thermal stress in the RCS pressure boundary.

Development of a Dynamic Event Tree for a Pressurized Water Reactor Steam Generator Tube Rupture Event

The DPRA-SGTR computer program was written to develop a dynamic event tree for the analysis of a steam generator (SG) tube rupture (SGTR) event. Using the dynamic event tree, a full-scope understanding of the possible responses of a plant following an SGTR event and the related actions with the emergency operating procedures (EOPs) can be analyzed. RELAP5/MOD3.2 was linked to DPRA-SGTR to calculate the thermal-hydraulic response of a Westinghouse three-loop pressurized water reactor at the Maanshan nuclear power plant. One SG tube with a double-ended break was postulated at the beginning of the accident. The plant thermal-hydraulic behaviors, status of the mitigation systems, and operator actions following the EOPs were explicitly modeled in the postulated SGTR. A total of 131 sequences were generated after an SGTR event. Among the 131 sequences, 91 sequences with a frequency sum of 8.5 × 10–6 were stopped either because of low-occurrence frequency (<1 × 10–12) or because the preset mission time was reached (30 000 s after initiating the event). Seven out of the 91 sequences with a frequency sum of 6 × 10–9 were intentionally stopped as a fatal error occurred when RELAP5 was calculating the thermal-hydraulic response.