Steam Generator Tube Rupture Research Articles

Numerical simulation of boiling dynamics in liquid Lead-Bismuth injected with multi-rupture high-pressure water jets

Once a steam generator tube rupture (SGTR) accident occurs in a lead-based fast reactor, it can trigger a domino effect of heat transfer tube ruptures, leading to a loss-of-flow accident involving multiple ruptures. This paper employs numerical methods to investigate the jet boiling phenomenon associated with the injection of sub-cooled water into high-temperature lead‑bismuth eutectic (LBE) through multiple nozzles. It explores how nozzle spacing, number, and diameter impact jet characteristics. Additionally, it analyzes flow characteristics and heat transfer mechanisms from the perspective of jet interaction. The study identifies critical areas within the multi-nozzle jet, specifically the three-phase mixing region, LBE entrainment region, water core region, and steam plume region. In scenarios where nozzle spacing is narrow or nozzle diameters differ significantly, the jet's base may undergo fragmentation. The jet process is categorized into two stages: the transition stage and the stable stage. During the stable stage, the mass flow rate of sub-cooled water remains relatively constant, with nozzle spacing exerting minimal influence on it. When considering the cross-sectional area of a single nozzle, an increase in nozzle diameter elevates the normalized mass flow rate, whereas the number and diameter of nozzles have little effect on it. Jet penetration depth is inversely related to nozzle spacing. As the number and diameter of nozzles increase, penetration depth initially rises and then decreases. Nozzle spacing has a minor influence on the jet's steam volume, while the number and diameter of nozzles are positively correlated with steam volume. Conversely, the number and diameter of nozzles negatively correlate with normalized steam volume.

Research on Assessment Method for Cognitive Influence Factors of Nuclear Power Plant Operators Based on Fuzzy Analytic Hierarchy Process and Deep Learning of Electroencephalogram Signals

This study explores the factors influencing the cognitive processes of operators in digital nuclear power plants, with a focus on the correlation between these factors and electroencephalogram (EEG) features. Initially, based on expert consultations, seven factors were considered: stress, time, fatigue, procedural complexity, user interface experience, procedural clarity, and efficiency. From these, four were identified as the most crucial for each stage of the cognitive process, highlighting their significant roles in influencing cognitive performance and potentially correlating with distinct EEG characteristics. These were assessed using the fuzzy analytic hierarchy process (FAHP) to determine the weightings of influences across the cognitive stages of monitoring, decision making, and execution. Employing a simulated scenario of a steam generator tube rupture, subjective questionnaires were utilized to gauge participant perceptions of influencer impacts at each stage, calculating human factors fuzzy synthetic values. Concurrently, EEG signals were segmented by operational steps, extracting around 114 features across the time, frequency, and time-frequency domains, which were then dimensionally reduced to 17 principal components via adaptive principal components analysis (APCA). A correlation analysis was performed between the human factors fuzzy synthetic values and the APCA-reduced EEG features of participants. Subsequently, the EEG feature columns of the eight selected participants were used as inputs to construct a transformer-based self-attention network model to evaluate the participants’ human factors fuzzy comprehensive values. The findings confirm the transformer model’s efficacy in assessing these values, evidencing a significant correlation between the EEG features and human factors fuzzy synthetic values. Integrating FAHP with machine learning methodologies, this model proficiently estimated operators’ cognitive states during various cognitive processes, significantly enhancing human-machine interface design and the operational safety and efficiency at nuclear power plants.

Numerical simulation of the dynamic interaction characteristics between Wood’s metal and water under coolant injection mode

The accident of the steam generator tube rupture in the lead-bismuth-cooled fast reactor or the core meltdown in the light water reactor may cause violent heat and mass transfer between the melt and the coolant, potentially further damaging the reactor. Therefore, it is essential to study the corresponding mechanisms. In this study, numerical simulations are utilized to investigate the dynamic interactions between Wood’s metal and subcooled water in the coolant injection (CI) mode, focusing on jet behavior and critical cavity parameter changes. By analyzing these factors, this research aims to clarify the interfacial evolution and interaction mechanisms between the melt, water, and air, thereby tackling the problem of inadequate observation in experiments. A user-defined function (UDF) was employed to simulate the flow characteristics of the melt during the solidification process. This approach effectively addressed the limitation of Fluent’s inherent solidification model, where the fluid no longer participates in turbulent velocity calculations after solidification. In addition, an equivalence method was proposed according to the characteristics of two-dimensional simulations in Fluent. Comparisons with Cheng’s experimental results show that the equivalent model can accurately predict jet penetration characteristics while significantly reducing the demand for computational resources. According to this equivalence method, this research developed a new correlation for predicting the maximum jet penetration depth in the CI mode. The above results provide new perspectives for understanding the dynamic interactions between metals and coolants, which is significant for the application and optimization of technologies in the nuclear field.

Separate effect experimental test and theoretical investigate of pressure wave propagation and pressure rising process upon SGTR accident in advanced LFRs

Steam Generator Tube Rupture accident (SGTR) accident is one of the main safety events to the industrial application of Lead-cooled Fast Reactors (LFRs). This paper presents a separate effect test facility, called Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST), for testing high-pressure subcooled water jetting into a Lead Bismuth Eutectic (LBE) pool, and analyzes experimental data of maximum initial pressure peak, dynamic pressure attenuation together with the pressure rising process upon SGTR. The experimental analyses put forward a general equation form together with a dimensionless number of the initial pressure peak associated with a high-pressure Newtonian fluid injecting into another low-pressure Newtonian fluid, and the derivation of the general equation form is specifically applied to match a further semi-empirical equation for an initial maximum pressure peak with 5–10 MPa water injection in SGTR. The attenuation curve of dynamic pressure wave in each experiment is matched, and more weakening conditions can change the pressure peak from concentrated and non-period peak group to approximate period 1.4–3.7 s pulse group. The size relationship between pipe length L and nozzle diameter D, which L/D=12, is set to divide SGTR into small wall rupture and tube shear fracture, further three stages of pressure rising are distinguished in the tube shear fracture. This work provides experimental data to help determining the severity of SGTR and assessing the consequences of this event as part of the safety case for the industrial application of LFRs.

A cumulative damage model for tubes in nuclear power plants subject to continuous degradation and multi-effect shocks under dynamic environments

Continuous degradation and shocks are two vital damage mechanisms for tubes under dynamic environments in nuclear power plants (NPPs). In this work, we presented a cumulative damage model to evaluate the reliability of tubes in NPPs, which accounts for continuous degradation and random shocks. The dynamic environments were assumed to obey a homogenous Markov process. Continuous degradation was described as a general stochastic process with independent increments. A multinomial distribution was used to depict the multiple effects of Poisson-distributed shocks on tubes. The influence of dynamic environments on the cumulative damage process of tubes was also explicitly modeled. Closed-form reliability measures were derived, such as reliability functions and mean time to failure (MTTF) of tubes. Drawing upon a case study from the existing literature, the analytical solution was validated through a comparative analysis with the outcomes of Monte Carlo simulation (MCS) and other methodologies referenced in the literature. As a practical engineering application, the reliability of steam generator tubes made of Alloy 690 within pressurized water reactors (PWRs), which are susceptible to primary side stress corrosion cracking (PWSCC) and the effects of water hammer, was calculated. The findings indicate that the probabilities of steam generator tube rupture (SGTR) as estimated by the model and those inferred from operational experience have the same order of magnitude.

A case study to address the limitation of accident scenario identifications with respect to diverse manual responses

Probabilistic safety assessment (PSA) is a tool for securing the operational safety of nuclear power plants. One unique benefit of PSA is to identify accident scenarios leading to undesirable consequences. Since these accident scenarios result from highly complicated combinations of automatic and manual responses, a huge number of simulations are generally required with a precise thermal-hydraulic (TH) code. Therefore, current PSA typically reduces the number of TH code runs by focusing on limited numbers of safety-critical components with reasonable engineering assumptions. Nevertheless, it has been pointed out that the current catalog of accident scenarios in PSA may be insufficient to some extent. In order to corroborate this claim, in this work, 14,348,907 simulation conditions were randomly generated after combining 79 manual responses identified from the emergency operating procedure of a steam generator tube rupture (SGTR) event in a reference nuclear power plant. Results show that about 14.7 % of the simulations correspond to an undesirable consequence pertaining to the SGTR. Accordingly, in terms of enhancing the quality of PSA results, the results of this study support that it is inevitable to develop a promising way to reduce the amount of computational resources required by large numbers of TH code runs.

Numerical simulation and model development of drag coefficient of bubbles in gas-liquid metal two-phase flow

The occurrence of a steam generator tube rupture (SGTR) accident in a lead-bismuth cooled fast reactor results in the formation of steam bubbles in the liquid lead-bismuth eutectic (LBE). This may degrade heat transfer and power transients in the reactor core due to the migration and accumulation of steam bubbles. To investigate the dynamics of steam bubbles flowing in liquid LBE, it is essential to develop an accurate model for the bubble drag coefficient. In this paper, a three-dimensional numerical model is first established to simulate the injection of high-pressure steam bubbles into a high-temperature LBE molten pool. The model is based on the CLSVOF method. By analyzing the trajectory, velocity, and diameter of bubbles, and combining them with the force equilibrium equation for bubbles, the values of the drag coefficient for bubbles are determined. On this basis, the suitability of current empirical drag models for bubble migration in LBE is evaluated. Finally, the optimal drag coefficient model is selected and further improved. Results reveal that the prediction error of the optimized model for the bubble drag coefficient in liquid LBE is within ±15 %.

Assessment of Source Terms and Potential Doses Due to Steam Generator Tube Rupture of VVER-1200 at the El Dabaa Nuclear Power Plant in Egypt

Assessment of Source Terms and Potential Doses Due to Steam Generator Tube Rupture of VVER-1200 at the El Dabaa Nuclear Power Plant in Egypt

Numerical analysis and structural optimization of a DLC coated capacitance wire-mesh sensor for measuring liquid metal-gas two-phase distribution

The detection of liquid metal-gas two-phase distribution in lead-cooled fast reactor once steam generator tube rupture occurs is vital to the security assessment of the reactor core. In this paper, considering the high temperature, corrosiveness and high conductivity of the liquid metal, a diamond-like carbon coated capacitance wire-mesh sensor is proposed to detect the two-phase distribution in liquid metal. Based on the software COMSOL, a numerical model is firstly established that is quantificationally verified by experiment to study and optimize the performance of the coated WMS. The measurement accuracy of the sensor is investigated by ranging the AC excitation frequency and structure parameters of the system. The results show that the liquid metal has blocking effect on the small bubbles and the sensor has great measurement accuracy when the transverse and vertical interval are 2 mm while the exitation frequecy, coat thickness and electrode diameter show little effect. This research could guide the measurement in liquid metal-gas two-phase flow.

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.

Full scope scaling analysis and recommendations for PWR SGTR scenarios

Scaling studies were conducted in the field of nuclear thermal-hydraulics to analyze the scaling distortions in a Steam Generator Tube Rupture (SGTR) scenario. Test 4 of the NEA/OECD/ROSA-2 Large-Scale Test Facility (LSTF) was chosen as the basis for the analysis of scaling distortions. Firstly, a post test calculation of the experiment was performed with the RELAP5 system code along with a phenomenological assessment. This model was then scaled over a wide range of sizes to investigate the possible scaling distortions by means of the Power-to-Volume Scaling Tool (PVST) which is an in-house tool developed and validated by the Universitat Politècnica de Catalunya (UPC) for scaling nodalizations. The paper demonstrates that the main scaling distortion observed in Test 4 is attributed to the lack of preservation of the L/D ratio in the SGTR assembly nozzle. This rationale is quite important in the design of break units in experimental facilities to preserve the friction and the mass flows of the reference (prototype) plant and scenario. Subsequently, the paper presents a series of guidelines for minimizing scaling distortions and reports the results obtained by applying them, the proportional scaling distortion was reduced by more than tenfold. Additionally, the paper analyzes the RELAP5 models involved in the conservation of momentum equation for the SGTR assembly nozzle, and their potential scaling distortions.

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.

Iodine source term assessment under DBA SGTR accident scenario

In the framework of the European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA), application of Best Estimate advanced codes for evaluating the radiological release under Design Basis Accident and Design Extended Conditions was considered. The aim is to assess the current modelling capabilities and to propose new calculation methodologies in order to produce more realistic evaluations of radioactive releases resulting from such accidents.In this paper, the studies are conducted for a reference Design Basis Accident Steam Generator Tube Rupture scenario in a 3-loop PWR-1000 reactor in terms of key thermal–hydraulic and radiological variables that condition and characterize the fission product release to the environment. The assessment includes a comparative study of the results performed by three organisations using three different computational tools and approaches. The outcomes highlighted issues considered for the design basis evaluation, modelling differences as well as some challenges and limitations in carrying out such analysis.

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.

Methods for the radioactive release estimation under DEC-A conditions

After the Fukushima Daiichi Nuclear Power Plant accident, the importance was raised to strengthen the assessment of the safety level of Nuclear Power Plants (NPP) by considering situations more severe than those integrated during the design of the plants. Design Extension Conditions (DEC) term was introduced by IAEA and WENRA. One of the most important subjects in the analysis of these conditions is the evaluation of radiological consequences.The Reduction of Radiological Accident Consequences (R2CA) collaborative project started in 2019 in the frame of the Horizon-2020 Program of the European Commission. The project addresses a broad scope of LWR designs (Gen II, III and III +) through the analyses of bounding scenarios of Loss Of Coolant Accidents (LOCA) and Steam Generator Tube Rupture (SGTR) transients and explores DBA and DEC-A conditions. The Lithuanian Energy Institute is taking part in this project as well.The part of work provided in the frame of R2CA project is presented in this article. The generic BWR-4 type reactor with MARK-I containment was analyzed in the case of DEC-A conditions. In this analysis the main thermal hydraulic processes, fission product release and its transport from the broken loop through containment to the environment were investigated. The number of ruptured fuel assemblies is the key parameter determining the fission product release from the core. The methodology which would help more precisely evaluate the number of failed fuel assemblies is presented in this work. The methodology consists of the application of integral severe accident code ASTEC and fuel performance code TRANSURANUS. Analysis using ASTEC code showed the importance of the reactor core nodalisation scheme. Taking iterative calculations, the relative power to which fuel assemblies remain intact at the selected DEC-A condition scenario was obtained. These results were verified with TRANSURANUS code calculations considering the uncertainties of the most relevant parameters. Based on TRANSURANUS calculation results, the ASTEC core nodalisation was improved. Improved ASTEC model nodalisation allowed to calculate more realistic rates of fission products releases from the fuel to the coolant, from the coolant to the containment and as well from the containment to the environment.