Steam Generator Tube Rupture Accident Research Articles

Evaluations of radionuclide activity releases into environment during loss of coolant accidents using the ASTEC code in pressurized water reactors within design basis and design extension conditions

The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project.More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project.This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios.

Design and optimization analysis of a new double-layer tube type heat exchanger for lead-bismuth reactors

Double-layer heat tubes have been designed to effectively reduce the occurrence of heat pipe rupture accidents. However, inter-tube thermal contact resistance can decrease heat transfer efficiency, thus hampering the heat dissipation in the primary loop system of lead-bismuth reactors. Therefore, optimizing the design of double-layer heat tubes is necessary. This work focuses on the double-layer heat exchanger of a lead-bismuth reactor and utilizes gallium-based graphene nanofluids as a thermal interface material to fill the gap between the heat tubes. Furthermore, the impact of the length, wall thickness, outer diameter, and spacing of heat tubes on the heat transfer performance of the double-layer heat exchanger with and without the nanofluids has been analyzed. The study aims to optimize the JF factor and cost-effectiveness ratio (CER). Genetic algorithms are employed to optimize and evaluate the heat transfer performance of the main heat exchanger based on the four aforementioned parameters. Consequently, a new design scheme is obtained for the double-layer heat exchanger, which increases the optimized overall heat transfer coefficient of the main heat exchanger by 5.79%, pressure drop in the primary loop by 2.32%, JF factor by 5%, and CER by 24.62%. These results demonstrate that the gallium-based graphene nanofluids can effectively enhance the heat transfer performance of the double-layer heat exchanger while reducing the likelihood of steam generator tube rupture accidents.

Safety evaluation of Multiple Steam Generator Tube rupture accident using the best estimate plus uncertainty approach

Following the disaster in Fukushima Nuclear Power Plant (NPP) in 2011, the awareness of securing the safety of NPP under extreme events has been raised. For this purpose, the concept of Design Extension Conditions (DECs) was introduced, to enhance the plant’s capability to withstand events that are more severe than Design Basis Accidents (DBA), such as Multiple Steam Generator Tube Rupture (MSGTR) accident, given its rapid progression and significant potential for large release of radioactive materials to the environment. It is worth noting that MSGTR has never occurred during the commercial operation of nuclear power plant, therefore the knowledge of consequences of the event and potential mitigation strategy is limited. In this study, a postulated MSGTR (simultaneous rupture of 5 U-tubes in a single SG) accident is analysed using the best-estimate thermal hydraulics code RELAP5/SCDAPSIM/MOD3.4. The main focus of this work is evaluation of appropriate operator mitigation actions under various inherent uncertainties. To this end, the best estimate plus uncertainty (BEPU) approach is adopted, whereby RELAP5/SCDAPSIM/MOD3.4 was coupled to the statistical analysis software, DAKOTA, to conduct the uncertainty quantification (UQ). Using BEPU analysis, an ensemble of system responses under various epistemic and aleatory uncertainties is obtained and used to validate the efficacy of operator actions in bringing the plant to a safe shutdown condition.

Iodine source term assessment as result of iodine spiking and mass transfer phenomena during a SGTR transient using MELCOR 2.2 and CATHARE 2 codes

The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies.The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools.The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MWe PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters.The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE.The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail.

Conjugate heat transfer characteristics numerical simulation on steam generator tube of lead-cooled fast reactor

The safety assessment of the Steam Generator Tube Rupture (SGTR) accident is a crucial step in the design process of the Lead-cooled Fast Reactor (LFR). To begin with, it is necessary to simulate the steady-state heat transfer of the LFR Heat Exchanger (HX) in order to determine the initial conditions of the SGTR accident. In this study, the heat transfer characteristics of the fluids on both sides of the LFR HX are analyzed. The appropriate models are modified based on the heat transfer characteristics of liquid metal and boiling water. The accuracy of the boiling models is verified by comparing the simulation results with the outcomes of boiling heat transfer experiments. Additionally, the simulation results are compared with the results obtained from ATHLET calculation and empirical formulas. On the secondary side, the supercooled water undergoes heating until it reaches the superheated steam phase. The preheating section spans from 0 to 600 mm, the saturated boiling section ranges from 600 to 3000 mm, and the superheating section extends from 3000 to 6600 mm. Ultimately, the distributions of steady-state temperature, pressure, and vapor volume fractions of the LFR HX are determined, providing the necessary initial conditions for the occurrence of the SGTR accident.

Performance evaluation of an improved pool scrubbing system for thermally-induced steam generator tube rupture accident in OPR1000

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

Bubble fragmentation characteristics during the injection of subcooled water jet into hot liquid pool

Bubble fragmentation and size distribution in hot and cold fluid jet interaction are fundamental for industries such as energy and metallurgy, especially for safety analysis after steam generator tube rupture (SGTR) accidents in lead-cooled fast reactors. The opaqueness to light of the molten lead pool makes it difficult to capture the gas-liquid interface behavior and thus lacks relevant data. In this paper, silicone oil was used instead of lead to simulate the injection process of the SGTR accident. Then the jet visualization experiments were carried out to investigate the bubble fragmentation characteristics. The visualization results show that the jet between the hot and cold fluids process is accompanied by strong thermal interactions, eventually leading to many bubbles dispersed in the high-temperature liquid pool. Secondly, four mechanisms of action affecting bubble size were elaborated in this paper, among which local vapor explosion is the primary mechanism of bubble fragmentation. Finally, the data revealed that bubble size presented a lognormal distribution, and the fragmented bubble size distribution was modeled. This study is promising to provide guidance for the bubble behavior characteristics after SGTR accidents in lead-cooled fast reactors.

Bubble transport during SGTR accident in lead-cooled fast reactor: A machine learning

Steam generator tube rupture (SGTR) is one of the safety issues for pool-type lead-cooled fast reactors (LFR). Accurately quantifying and predicting the bubble transport result is essential for evaluating the accident. This paper addresses tracking the bubble motion using an Eulerian-Lagrangian method in CFD based on the Europe Lead cooling System (ELSY) primary system model at 1/8 centrosymmetric structure. The steady and transient bubble distributions in the system under different leakage heights are obtained. Furthermore, the simulation results are predicted by machine learning, where Gaussian Process Regression (GPR) is employed for both steady conditions and transient conditions. The data-driven GPR model is established for the bubble transport prediction, which can capture complex non-linear relationships between input variables (i.e. bubble diameter, SG leakage height, final position, system contaminate) and output (reached bubble percentage) responses directly from data. For steady conditions, the prediction results by the kernel function of Automatic Relevance Determination (ARD) Rational Quadratic show the best accuracy in predicting the percentages of bubbles reaching the core, top of the steam generator, and staying in the system, with a total Root Mean Square Error (RMSE) of 3.22%. Four typical transient cases of bubble accumulation in the core are selected for prediction. All the cases are well predicted by the kernel function of ARD Matern 3/2 with low mean averaged error.

Multiphase simulation of hyperbaric steam-water jet inside liquid Pb-Bi eutectic environment

Numerical simulation analyses of LIFUS5/MOD2 experiment facility for hyperbaric steam-water jet inside liquid lead bismuth eutectic (LBE) were conducted using computational fluid dynamics (CFD). Alchagirov surface tension correlation and Lee evaporation model were applied in Eulerian-Eulerian two-phase flow model. Benchmark study was performed by comparing the CFD results against experimental data. Four thermal-hydraulics phenomena of sensitivity analysis were established under three extended conditions. These include 1) argon zone compression risk, 2) pressure rising and initial pressure peak, 3) argon entrained by LBE fluctuation, and 4) LBE level rising and steam cavity diffusion path. Overall, CFD models have a great advantage simulating pressure change and pressure wave transfer with 0.37% maximum error. LBE level rising and the radial and even diffusion of steam cavity should happen simultaneously under steam generator tube rupture accident (SGTR). This study provides the engineering basis for Lead-cooled Fast Reactor (LFR) design to improve its safety margin.

CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam

CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam

Coupled modeling of neutronics and thermal-hydraulics processes in LFR under SG-leakage condition

The lead cooled reactor BREST-OD-300 is developing as a part of Russian federal project “PRORYV”. Two- circuit scheme is used in the reactor for heat removal. An inherent risk of two- circuit reactor is the potential danger of water steam ingression in the core in the case of large leakage in steam generator initiated, for example, Steam Generator Tube Rupture (SGTR). Reactor power and temperature response on vapor penetration to the core is studied, but pressurization effects are not in the purview of the paper. The 3D multi-physics (neutronics + thermal-hydraulics) UNICO-2F code was developed for study of SGTR accident. The code calculates unsteady 3D space dependent distributions of coolant velocity, pressure and temperature, space distributions of vapor concentration and heat release density in the core and 3D temperature distributions in the fuel pins. Guillotine rupture of one tube in Steam Generator (SG) is considered as initial event of the accident. It is shown that even with the most conservative assumptions reactivity insertion due to vapor ingress in the core causes small increase of power in level and as a result maximum cladding temperature continue to stay well below safe operation design limit in the entire transient. Hypothetical option of simultaneous tube rupture in few SG belonging to different loops is also analyzed. It is demonstrated that even in the case of simultaneous large leak in two SG the transient stays mild and temperature in the core after two small oscillations is stabilized at acceptable level. In the long term the analysis confirmed the high level of reactor self-protection against SGTR accident.

Study on water jet penetration behavior in molten LBE during SGTR accident with simulant experiments

During Steam Generator Tube Rupture (SGTR) accidents of Lead-bismuth cooled Fast Reactor (LFR), the coolant water in the secondary loop will directly contact with the high temperature molten lead–bismuth eutectic (LBE) in the primary loop around the rupture point resulting in the strong interaction, which may lead to vapor explosion, and eventually damage the structural components of the reactors and cause the possible leakage of radioactive material. In order to further investigate the penetration behavior of water jets in molten LBE, an experimental platform for visualizing the two-phase flow of liquid jets is designed and built, and a series of simulant experiments are conducted by injecting water or ethanol jets into fluorinert FC-40 under different experimental conditions. By analyzing the penetration process of ethanol jet in FC-40, the flow pattern evolution and penetration law of ethanol jet under boiling and non-boiling conditions are obtained. The boiling effect is considered to be the key factor affecting the penetration depth of jet in the range of current experimental conditions. By comparing the injection behaviors of ethanol and water, the consistent flow pattern shows that the injection behavior can be described with the same relation. The results of dimensional analysis method for jet penetration process show that the Froude number theory can reasonably predict the maximum dimensionless penetration depth under different jet conditions.

Using machine learning to forecast and assess the uncertainty in the response of a typical PWR undergoing a steam generator tube rupture accident

In this work, a multivariate time-series machine learning meta-model is developed to predict the transient response of a typical nuclear power plant (NPP) undergoing a steam generator tube rupture (SGTR). The model employs Recurrent Neural Networks (RNNs), including the Long Short-Term Memory (LSTM), Gated Recurrent Unit (GRU), and a hybrid CNN-LSTM model. To address the uncertainty inherent in such predictions, a Bayesian Neural Network (BNN) was implemented. The models were trained using a database generated by the Best Estimate Plus Uncertainty (BEPU) methodology; coupling the thermal hydraulics code, RELAP5/SCDAP/MOD3.4 to the statistical tool, DAKOTA, to predict the variation in system response under various operational and phenomenological uncertainties. The RNN models successfully captures the underlying characteristics of the data with reasonable accuracy, and the BNN-LSTM approach offers an additional layer of insight into the level of uncertainty associated with the predictions. The results demonstrate that LSTM outperforms GRU, while the hybrid CNN-LSTM model is computationally the most efficient. This study aims to gain a better understanding of the capabilities and limitations of machine learning models in the context of nuclear safety. By expanding the application of ML models to more severe accident scenarios, where operators are under extreme stress and prone to errors, ML models can provide valuable support and act as expert systems to assist in decision-making while minimizing the chances of human error.

Comparative study of data-driven and model-driven approaches in prediction of nuclear power plants operating parameters

Nuclear energy plays a crucial role in mitigating climate change as a near carbon-free and clean energy source. Predicting operating parameters is key to the digitalization and intellectualization of nuclear power plants, improving energy efficiency and reducing costs. Parameter prediction methods mainly consist of model-driven and data-driven approaches, and a comparative study is necessary to select the appropriate prediction method or combine them. In this paper, the gated recurrent unit network and the thermal–hydraulic program RELAP5 are chosen as representative data-driven and model-driven approaches to be used for predicting parameters for a specific operating condition to assess the characteristics and capabilities of each. The chosen experiment is a steam generator tube rupture accident which is important to pressurized water reactors. In this paper the modelling process and the prediction results for both approaches are given. The correlation coefficient for dome pressure, dome temperature, and primary outlet temperature of both approaches is greater than 0.993, except that the correlation coefficient of downcomer temperature is 0.83865 using RELAP5 and 0.99392 using gated recurrent unit network. Then the two approaches are compared from six perspectives of practical application, including accuracy (gated recurrent unit network and RELAP5 received scores of 8 and 6, respectively.), data requirements (7 and 5, respectively), model building algorithm (7 and 5, respectively), prediction speed (10 and 6, respectively), expertise requirements (6 and 6, respectively), and interpretability (5 and 9, respectively). It shows that model-driven and data-driven approaches both require specialist knowledge and appropriate data for modelling process. The data-driven gated recurrent unit network exhibits better accuracy and higher speed. The model-driven RELAP5 has better interpretability compared to the black-box gated recurrent unit network. It is suggested that hybrid approaches of model-driven and data-driven approaches could better deal with the prediction problems.

Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration

One of the main safety issues of Generation IV (Gen IV) heavy liquid-metal fast reactors is the postulated steam generator tube rupture (SGTR) accident. This event is characterized by primary and secondary coolant interaction, referred to in the literature as a coolant-coolant interaction event having a nonzero probability to occur. This accident scenario could affect the safety of a pool-type reactor, as a consequence of water secondary coolant flashing into the primary coolant liquid metal. The SGTR event needs to be experimentally characterized to evaluate the pressure waves effect, tube rupture propagation (domino effect), oxide precipitation and slug and plug formation, cover gas pressurization of the reactor, and steam flow paths through the pool and eventually the core, entailing the risk of positive reactivity insertion (due to positive local void coefficient). The design phase of the Gen IV MYRRHA plant has dealt with postulated SGTR safety issues in the framework of the MAXSIMA project, which is supported by the European Commission. A relevant contribution to this research activity was provided by the Italian Agency for New Technologies, Energy and Sustainable Economic Development Research Center Brasimone, where a new test section has been designed, assembled, instrumented, and implemented in the large-scale integral-effects pool facility CIRCE for investigating the SGTR event in a relevant configuration for the heat removal system of MYRRHA. This research reactor is not oriented to steam production for running a turbogenerator (no electric production), thus the heat removal system is referred to as the primary heat exchanger (PHX) and not as a steam generator. Four full-scale portions (four bundles of 31 tubes) of the MYRRHA PHX were adopted to carry out four independent SGTR experiments. Water flowed upward in the central tube of the bundle and two rupture positions were investigated at the lower and upper levels, named the bottom and middle scenarios, respectively. After the rupture, water was injected at 16 bar and 200°C into lead bismuth eutectic alloy at 350°C. The experimental results showed a remarkable repeatability and were presented in terms of (1) CIRCE vessel pressurization up to 2.7 and 4 bar absolute for the middle and bottom scenarios, respectively; (2) vapor flow paths through the bundle and its cooling effect up to 120°C and 140°C for the middle and bottom tests, respectively; and (3) strain measurements on tubes and bundle shells up to 2800 μm/m. The integrity of the tubes surrounding the ruptured one and the effectiveness of implemented safety device (rupture disks) pressure relief were significant engineering feedback for MYRRHA designers. The acquired high-quality data also constitute a database increase for future code verification and validation and possible new model development. The performed experimental analysis provided the awareness that a suitable design of a depressurization system (e.g., rupture disks) could allow for addressing postulated SGTR events in the MYRRHA configuration with confidence and safety.

Study on the accidents analyses of a single channel for XADS by using MPC-LBE code

Abstract Accelerator Driven sub-critical System (ADS), which employs the high-energy proton beam generated by accelerator to bombard the target nucleus and generate spallation neutrons as external neutrons to drive and maintain the operation of its sub-critical reactor, is of great significance in nuclear waste treatment and disposal. As the instability of proton beam would affect the power level of the reactor and threaten the safety of ADS, Beam Trip (BT) and Beam OverPower (BOP) are commonly considered to be its two typical transient accidents. As for the sub-critical reactor, the Transient OverPower (TOP) is also one of typical transient accidents that should be considered, which is mainly caused by reactivity insertion under certain cases, such as SGTR (Steam Generator Tube Rupture) accident. For the subcritical reactors, the transient evolution behaviors are strongly affected by the subcriticality value. On the one hand, the subcriticality values of ADS design should take safety margin and power gain into consideration. On the other hand, the subcriticality value is variable with the burnup of reactors. So it is necessary to study the safety characteristics of the subcritical reactors under different subcriticality values, in this paper, the transient safety characteristics of a single channel for XADS under BT, BOP and TOP accidents of different subcriticality values were investigated by using MPC-LBE code.

Experimental investigation of pressure fluctuation and steam transport under SGTR accident in Lead-cooled Fast Reactors

Transient multiphase flow phenomenon of hyperbaric water jetting into lead-based eutectic (LBE) is of significant importance for safety analysis of Lead-cooled Fast Reactors (LFRs). In order to investigate the thermal-hydraulic parameters under Steam Generator Tube Rupture (SGTR) Accident in LFRs, the Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST) was built and the effects of (1) initial pressure peak, (2) pressure rising, (3) steam cavity migration in reaction vessel were evaluated. Under the present experimental and theoretical analysis, the results showed that the maximum pressure peak of 2.93 MPa occurred under the condition of 10 MPa water initial pressure which is the highest experimental requirements of reactor second system, and the accident consequence of pressure peak together with the maximum rising rate 2.001 MPa/s might be the most serious. Floating rate of steam cavity was different in radial direction, and was affected by water pressure and LBE level height. The analysis results will provide severity assessment for safety design, accident forecasting and engineering guidance of LFRs.

CFD investigation on thermal hydraulics of the double-wall bayonet tube heat exchanger in CLEAR-S facility

The double-wall bayonet tube type heat exchanger design can monitor and mitigate the occurrence of Steam Generator Tube Rupture (SGTR) accidents, which is one of the most important accidents of GEN-Ⅳ lead-cooled Fast Reactor (LFR). The lead-based reactor engineering validation facility CLEAR-S will be used for the integrated testing and verification of the megawatt-level double-wall bayonet tube heat exchanger. In this study, the Shear Stress Transport turbulence model (SSTk-ω) was used to pre-simulate the thermal characteristics of the CLEAR-S double-wall bayonet tube heat exchanger before the experiment. The verification shows that the numerical simulation results of the single bayonet tube structure are in good agreement with the RELAP5 simulation results and experimental data from CLEAR-S. CFD analysis reveals that the non-effective heat exchange area will preheat the liquid water on the side of the tube, so that the overall heat exchange performance decreases by about 4%. The heat conduction of the powder interlayer is the largest heat transfer resistance, which weakened the influence of the lead-bismuth eutectic (LBE) side flow and temperature inhomogeneity on the heat transfer of the tube bundle and significantly reduced the wall radial temperature difference of each bayonet tube. The conclusions of this study can provide a reference for the experiment and design of double-wall bayonet tube heat exchangers.

Evaluation of aerosol retention inside a steam generator under steam generator tube rupture accident conditions

The necessity has arisen to conduct risk evaluations of the steam generator tube rupture (SGTR) accident for domestic operating nuclear power plants in the Republic of Korea in accordance with a recent amendment to the Nuclear Safety Act. To evaluate the aerosol-type radionuclides released to the environment during the accident and to prepare mitigation strategies, tests conducted in experimental facilities that faithfully reflect the properties of domestic plants are essential. In this work, one such experimental facility was installed at the Korea Atomic Energy Research Institute with scaling work considering the geometrical shapes of a domestic Korean plant and the major aerosol removal mechanisms inside the steam generators. A preliminary test was first conducted to understand the amount of aerosol removal inside the pipes of the facility with analytical methods. Then, aerosol retention tests were performed in both dry and flooded conditions. Decontamination factors were evaluated in each test, and the results were analyzed using filter measurements, an electrical low-pressure impactor, and deposition mass in the tubes. The test results are believed to provide a good starting point to improve the related accident management strategies to increase the safety of domestic Korean nuclear power plants.

Updating risk model for SGTR accident based on success criteria analysis

Abstract Success criteria analysis plays a key role in the development of realistic probabilistic safety/risk assessment (PSA/PRA) model because it provides supporting information regarding the response of complex nuclear power plant systems to different accident conditions. The current paper performs plant specific success criteria analysis for steam generator tube rupture (SGTR) accident in a typical pressurized water reactor (PWR) and demonstrates implementation of the obtained best estimate results on a risk model which has been initially developed based on expert judgment and general plant design data. The modifications on the risk model include configuration of the safety systems as well as the event tree structure. The updated PSA model shows 50% reduction in the plant core damage frequency (CDF) in comparison to the base case risk model. This highlights the importance of success criteria analysis for the development of a realistic PSA model in risk informed applications.