Cold Leg Research Articles

The thermal-fluid-solid coupling effect in nuclear reactor vessel with direct vessel injection

Different from conventional cold leg safety injection, a newly designed reactor type CAP1400 directly connects the injection pipeline of the emergency cooling system with the pressure vessel. However, these structures have been found to potentially have adverse effects on the pressure vessel. This paper employs both physical experiment and numerical analysis to investigate the dynamic oscillating load and the structural response. A 1:9 scale model is adopted for the experimental section. The uncertainty of the measurement quantities is within 3%. The CFD code and the FEM code are used to simulate the transient fluid flow and solid structure deformation, respectively. The sequential coupling method is used for the thermal-fluid-solid coupling calculation. The overall error of the calculation is within ±10%. The result shows that the fluctuation of the physical quantity near the wall is caused by turbulence eddy and thermal mixing, and the flow, heat transfer and structural deformation near the wall of the pressure vessel are closely coupled. The effect of the turbulence eddy can be seen within the entire range of frequency measured, while that of the thermal mixing can be seen only within the low frequency range. The influence law of the loop flow ratio, the flow rate of the downcomer and the loop temperature difference in the structural response are obtained. Through the research of this paper, the law of interaction between the thermal-fluid load and the structural response under direct safety injection is more deeply understood. Thus, it provides a design reference for the safety of reactors.

BETHSY 9.1b test calculation with TRACE using 3D vessel component

Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modeled and compared to 1D fluid dynamics. The TRACE 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. In general, all presented code calculations were in good agreement with the BETHSY 9.1b test. The TRACE 1D calculation results are comparable to RELAP5 calculated results. For some parameters they are better, this is mostly due to better tuning of the break flow, what influences timing of the transient. When comparing TRACE 1D and TRACE 3D calculation, the latter is slightly better. One reason for comparable results is already good agreement of 1D calculations and there was not much space to further improve the results. The other reason may be that in the facility the phenomena were mostly one dimensional (for example, external downcomer was used for reactor vessel modeling). However, when 3D behavior of the heater rod temperatures was investigated, the advantage of three dimensional treatment was clearly demonstrated.

Comparing Analysis of Loss of Coolant Accident on Bethsy Facility with Apros 6.05 and 6.06.

In this paper, the comparison of analysis of international standardized problem ISP 27 using two versions of APROS process simulation software 6.05 and 6.06 is presented. Numerical simulation of experiment Bethsy 9.1b, also known as ISP 27 was performed on a scaled down model of a three loop, 900 MWe Framatome PWR. In the test a small LOCA, with 2-inch cold leg break, combined with High pressure Injection System (HPIS) failure is simulated. State oriented approach, which requires operators to start an Ultimate Procedure were used. Model was first built in APROS 6.05 using standard modules in order to describe the volumes, heat structures and regulation of the test facility and was then exported to APROS 6.06. The results from both versions showed all the processes such as loop seal clearing, core uncover and rise of cladding temperature and other processes taking place in the experiment were in a good agreement with experimental data. However even though results were similar some differences were noticeable. The differences in core cladding temperature, time integrated break mass flow, core liquid level and pressurized pressure were analysed in more detail in this paper.

NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes

NPP Krško input deck developed at Faculty of Electrical Engineering and Computing (FER) Zagreb, for severe accident code MELCOR 1.8.6 is currently being tested. MELCOR is primarily used for the analyses of severe accidents including in-vessel and ex-vessel core melt progression as well as containment response under severe accident conditions. Accurate modelling of the plant thermal-hydraulic behaviour as well as engineering safety features, e.g., Emergency Core Cooling System, Auxiliary feedwater system and various containment systems (e.g., Passive Autocatalytic Recombiners, Fan Coolers and Containment spray) is necessary to correctly predict the plant response and operator actions. For MELCOR input data verification, the comparison of the results for small break (3 inch) cold leg Loss of Coolant Accident (LOCA) for NPP Krško using MELCOR 1.8.6 and RELAP5/MOD 3.3 was performed. A detailed RELAP5/MOD 3.3 model for NPP Krško has been developed at FER and it has been extensively used for accident and transient analyses. The RELAP5 model has been upgraded and improved along with the plant modernization in the year 2000. and after more recent plant modifications. The results of the steady state calculation (first 1000 seconds) for both MELCOR and RELAP5 were assessed against the referent plant data. In order to test all thermal-hydraulic aspects of developed MELCOR 1.8.6 model the accident was analysed, and comparison to the existing RELAP5 model was performed, with all engineering safety features available. After initial fast pressure drop and accumulator injection for both codes stable conditions were established with heat removal through the break and core inventory maintained by safety injection. Transient was simulated for 10000 seconds and overall good agreement between results obtained with both codes was found.

Krško NPP Experience with RTD Bypass Elimination

This paper describes Krško experience and problems that had to be resolved with RTD Bypass Elimination project (RTDBE). Following RTD bypass manifold isolation valve leak in 2008, Krško decided to perform RTDBE modification on reactor coolant system narrow range temperature measurement system. The installation was performed during Outage 2013. Soon after the plant returned to power, newly configured measurement channels showed that OPΔT reactor trip was oversensitive to spikes caused by auxiliary relay operation in the cabinets nearby. The solution was to reconfigure OPΔT trip filtering constants to filter out short-term spikes in the signal. After almost full operating cycle of trouble-free operation, RTD failures started occurring on reactor coolant system cold leg, which was caused by the high frequency vibrations (3-5 kHz) induced by reactor coolant pumps. To resolve RTD failures, Krško ordered re-design of the RTDs to add robustness and specific qualification in high-frequency vibration operating environment. Improved RTDs were installed in Outage 2016 and were operating one full cycle with minor deviations.

P50: Save leg: using contralateral femoral inflow for distal perfusion catheter placed for Impella CP

Background: Leg ischemia is a possible complication of percutaneous left ventricular device (Impella CP TM) placement. To avoid leg ischemia in atrisk patients, a distal perfusion catheter (DPC) should be placed. The DPC can be connected to the wire-reaccess port of the Impella CP as an inflow to the DPC to perfuse the leg. However, the wire-reaccess port of the Impella was not designed to bypass arterial blood to the distal limb, but rather for wire passage. We introduced a passive distal perfusion system from the contralateral femoral artery to optimize blood flow to the distal limb. Methods: A 65 year-old female with dilated cardiomyopathy developed refractory ventricular tachycardia. She was taken to the catherization laboratory because of hemodynamic instability requiring high doses of vasopressors. By left heart catheterization, we noted normal coronary arteries with elevated left ventricular end-diastolic pressure (38 mm Hg) and a poor cardiac index. Thus, Impella CP was placed from the right groin to stabilize her hemodynamics. A DCP was placed on right distal femoral artery and connected to the wire-reaccess port of the Impella CP. The leg appeared perfused by DPC via reaccess port in the catherization laboratory; however, despite this as well as appropriate anticoagulation, the leg became ischemic (cold leg, low leg regional saturation) shortly after admission to the ICU, where a lack of flow in the DPC was confirmed. Results: An additional contralateral femoral arterial line was placed. The contralateral femoral arterial line was then connected to the DPC (figure), and the reaccess port of the Impella was capped. Shortly after placement of the new DPC system, the pulse was returned and leg saturation improved. Additionally, leg ischemia during Impella support did not reoccur. Conclusion: A DPC using the Impella CP reaccess port provided inadequate distal flow. When a DPC is indicated during femoral Impella CP use, the contralateral femoral artery should be chosen for inflow cannulation. Figure 1. Impella CP and distal perfusion catheter (DPC) in the right femoral artery. The arterial line in the left femoral artery is feeding DPC in the right femroal artery.

Uncertainty quantification (UQ) for CFD simulation of OECD-NEA cold leg mixing benchmark

An uncertainty quantification (UQ) analysis was conducted for the unsteady Reynolds Average Navier Stokes Simulation (URANS) of the OECD-NEA cold leg mixing benchmark. The methodology is based on the American Society of Mechanical Engineers Verification and Validation (ASME V&V) standard for UQ in Computational Fluid Dynamics (CFD) applications. The numerical, input and experimental uncertainties were calculated to determine the validation uncertainty for the open test (T1). Furthermore, based on validation uncertainty, model uncertainty was calculated. Finally, on the basis of open test (T1) the symmetric and asymmetric uncertainty band was achieved for the blind test (T2). The validation uncertainty closely covers the PIV data in the cold leg. Besides, the simulation results of the blind test, together with the uncertainty band, closely bound the PIV data of the mean velocity in the cold leg. However, a large simulation uncertainty band was obtained in the downcomer due to a large validation error.

Comparison of Blind and Open Calculation Results for Top-Slot Break LOCA in Fourth ATLAS Domestic Standard Problem

The advanced thermal-hydraulic test loop for accident simulation (ATLAS) was developed and operated at The Korea Atomic Energy Research Institute. The ATLAS is operated to simulate accidents in pressurized water reactors (PWRs). A domestic standard problem (DSP) using the ATLAS was proposed for transferring the database of the integral effect test to Korean nuclear researchers and developing the safety analysis methodology of PWRs. The fourth DSP (DSP-04) exercise was performed during 2015–2017 with 15 participants (13 organizations), that are universities, government, and nuclear industries. In DSP-04, a top-slot break at the cold leg was chosen as the target scenario to resolve an issue about the effect of loop seal clearing and the reformation on the peak cladding temperature during the cold leg top-slot break LOCA for APR1400. The participants performed a code calculation for the experimental simulation and sensitivity studies for the enhancement of the code. This paper includes brief information about the experimental and major code assessment results.

Fractional order multi-scheduling parameters based LPV modelling and robust switching H∞ controllers design for steam dump system of nuclear power plant

In this research work, the highly challenging problem of novel modelling and nonlinear control of steam dump system of Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) is attempted. The Fractional Order Multi- Scheduling Parameters based Multi-Input Single- Output Linear Parameter Varying (FO-MSP-MISO-LPV) model of Steam Dump System (SDS) is estimated with uncertain dynamics under sudden load variation transients. MSP for uncertain dynamics of SDS in FO framework is the most challenging problem and attempted in a novel fashion for the first time in nuclear industry. Scheduling parameters are dynamic in nature that makes the control problem more challenging. The Model is estimated experimentally by least square method using innovative plant operational data of opening positions of different valves as input variables and steam pressure as an output variable and cold leg coolant temperature coefficient of reactivity, hot leg coolant temperature coefficient, steam flow rate and turbine power as dynamic scheduling parameters. A switching controller is designed to address variable conditions of steam pressure for the actuation of dump valves, relief valves and safety valves in SDS. A robust fractional order LPV switching H∞ (RFO-LPV-SWH∞) controllers are formulated and designed for FO-MSP-MISO-LPV model. The design of RFO-LPV-SWH∞ controllers is another significant contribution in switching mode with non-integer and LPV hybrid framework. RFO-LPV-SWH∞ controllers are tested, simulated and validated against benchmark transients as laid down in Final Safety Analysis Report (FSAR) of PWR-type NPP. The input and output variables at first and second vertex of polytope are fast reference tracking under highly nonlinear uncertain dynamics of SDS. Closed loop simulation experiments are conducted and proved that the proposed closed framework is robust in performance under parametric uncertainty.

Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes

The analyses of postulated severe accident sequences play a key role for the international nuclear technical scientific community for the study of the effect of possible actions to prevent significant core degradation and mitigate source term release. To simulate the complexity of phenomena involved in a severe accident, computational tools, known as severe accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2 ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a generic western three-loops PWR-900 MWe, has been carried out with the aim to give some insights on the modelling capabilities of these tools and to characterize the differences in the calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark exercise. In the postulated transient, the unavailability of all active injection coolant systems has been considered and only the injection of accumulators has been assumed as accident mitigation strategy.

Cold Leg LBLOCA uncertainty analysis using TRACE/DAKOTA coupling

Safety analyses for nuclear power plants were carried out in the past using a conservative approach. With the increase of the phenomenological knowledge, through experimental data, and computational power, it became possible to adopt best estimate thermal- hydraulic system codes to perform deterministic safety analyses. However, some uncertainties are still present in the models, correlations, initial and boundary conditions, etc. Therefore, it is fundamental to quantify the uncertainty of calculation. This approach is called “Best Estimate Plus Uncertainty” (BEPU). Among the available uncertainty analysis methodologies, the probabilistic method to propagate input uncertainty is widely adopted. In the present study, an uncertainty analysis of a cold leg large break loss of coolant accident in a generic PWR-900 MWe has been developed and it has been carried out coupling the best estimate thermal- hydraulic system code TRACE and the uncertainty quantification tool DAKOTA in the SNAP environment/architecture.

Numerical analysis on in-core ignition and subsequent flame propagation to containment in OPR1000 under loss of coolant accident

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium–concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

Fractional Order Neural Transient Modeling of Primary Circuit of ACP1000 Based Nuclear Power Plant in LabVIEW

The primary circuit of the nuclear power plant is the most advanced and sophisticated loop of the Advanced Chinese Pressurized Water Reactor (ACP1000). The primary circuit is composed of most technologically advanced nuclear systems and controllers. In this research work, closed loop dynamics of primary circuit (CLPC) of ACP1000 based nuclear power plant is identified. The closed loop dynamics is comprised of highly nonlinear coupled sevencontrol systems. The turbine power, pressurizer temperature, cold leg temperature, hot leg temperature, coolant average temperature and feed water flow are the selected parameters of interest as inputs while neutron power, reactor coolant pressure, pressurizer level, steam generator pressure, steam generator level and steam generator flow as outputs. Therefore, a closed loop multi-input multi-out (MIMO) is configured. The control oriented closed loop dynamics of the primary circuit of ACP1000 is estimated by state-of-the-art novel fractional order neural network (FO-ANN) tool developed in LabVIEW. The parameters of FO-ANN of CLPC (FO-ANN-CLPC) are optimized using fractional order backpropagation (FO-BP) algorithm. The performance of FO-ANN-CLPC is tested and validated in transient conditions and the proposed model predicted the desired reactor power with minimizing error function. The robust performance of the proposed closed loop model is evaluated by dynamic simulation for a prescribed turbine load power increase transient from 20 % to 100 % and validated against reactor power and behaviour of various thermal hydraulics parameters are observed and analyzed.

Analysis of a Hypothetical Complete Loss-of-Coolant Accident in a Typical Research Reactor

In the present study, an analysis of a hypothetical complete loss-of-coolant accident in a typical open-pool research reactor is conducted. The reactor core is assumed to be completely uncovered and exposed to the ambient air. The possibility of passively cooling the decay heat of the exposed reactor core by natural convection to air and thermal radiation until core reflooding is investigated. A three-dimensional computational fluid dynamics analysis is conducted for the uncovered core while cooled by air natural convection and thermal radiation. The reactor core is simulated as a porous zone with decay heat generation specified as a cosine-shape distribution. The reactor core decay heat acts as a driving force for the coolant flow from the cold leg to the hot leg. The thermal equilibrium porous media model is used to represent the energy equation inside the core region. This study is conducted for core uncover time (the time interval between reactor shutdown and the moment when the reactor core is drained of water) of 10E3, 10E4, 10E5, 10E6, 10E7, and 10E8 s. Contour plots of temperature, velocity, density, and pressure at different values of core uncover time are illustrated. It’s found that for core uncover times of 10E3, and 10E4 s, the maximum core temperature exceeds the cladding melting point. The core maximum temperature is well below the melting point for uncover times of 10E5, 10E6, 10E7, and 10E8 s.

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Abstract PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 8C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

Experimental investigation on thermal hydraulic interaction of RCS (reactor coolant system) and containment for intermediate break loss-of-coolant accident (IBLOCA) scenario in ATLAS-CUBE test facility

In order to experimentally investigate the thermal hydraulic interaction between the reactor coolant system (RCS) and the containment, the integral effect tests were conducted in a test facility, ATLAS-CUBE. Three tests for an intermediate-break loss-of-coolant accident (IBLOCA) were conducted for simulating the break of a cold leg or a direct vessel injection (DVI) nozzle. The effect of the mass/energy (M/E) supply from the break and the asymmetric phenomena inside the containment according to the break simulation position were investigated with considering its impact on the pressure/temperature (P/T) behavior inside the containment. As the test results, both of the RCS and the containment were effectively cooled down during the transient, by activation of the safety injection system and the containment spray system, respectively. The M/E supply from the RCS increased a pressure and a steam-gas mixture temperature of the containment, while a wall condensation on the passive heat sink and a spray injection contributed to cool down the steam-gas mixture. The two-phase flow characteristics of the break flow such as the quality or the subcooling of the liquid phase could highly affect the P/T build-up of the steam-gas mixture in the containment. The ATLAS-CUBE test data in this study will contribute to evaluate the thermal hydraulic safety analysis codes for the RCS and the containment.

ATHLET Simulation of PKL I2.2 IB-LOCA Benchmark Test and Quantitative Assessment

The Primary Coolant Loop Test Facility [Primӓrkreislӓufe Versuchsanlage (PKL)] PKL I2.2 Benchmark experiment for an intermediate-break loss-of-coolant accident (IB-LOCA) with a 13% or 17% break at the cold leg was performed in the Organisation for Economic Co-operation and Development/PKL-4 project at PKL in Erlangen, Germany. Analysis of Thermal-Hydraulics of LEaks and Transients (ATHLET) 3.1A was used at Karlsruhe Institute of Technology for its posttest calculations. Crucial predicted parameters were compared with measured data. The calculated trend of the selected parameters fits well with that of the experimental data except for the phenomenon of core heatup and the value of the peak cladding temperature. A fast Fourier transform–based method was chosen to quantify the matching of the parameter trends. According to the quantitative assessment, the IB-LOCA scenario and its detailed phenomena can be predicted well by ATHLET. Additionally, some discrepancies, i.e., insufficient reliable predictions for break mass flow and for reactor pressure vessel collapsed water level, were also observed, possibly deserving another study to undergo deeper scrutiny.

Experimental demonstration of the performance of the novel thermosyphon heat transport device and comparison with CFD predictions

Thermosyphon heat transport device (THTD) is a novel system operating on the principle of single-phase natural circulation driven by density difference and is significantly different from a heat pipe wherein the flow is pressure driven. The present experimental and numerical analyses deliberate the heat transportation capability of THTD and its flow phenomena, which are linked to the operating conditions viz. Power, heating length, and flow rate of coolant. Experiments confirmed the instability for the entire range of parameters tested as a chaotic oscillation of temperature rise across the heater and temperature drop across the cooler is seen. Additionally, the temperature rise across the heater was found to be significantly larger than the temperature drop across the cooler, indicating the flow is not purely unidirectional natural circulation flow, which was further confirmed by CFD simulations. It also revealed that the unstable flow pattern is a unidirectional flow with a recirculation loop in the cooler, which has both downward and upward flows. The extent of downward flow is found to depend on the operating conditions. Further, the unidirectional flow is observed in the entire cold leg, heated section, and near to the inner adiabatic wall along the hot leg and the cooler. The essence of the study is that despite the unstable nature of flow for the entire range of parameters tested, the heat transport capability of THTD is uninterrupted.

Safety Analysis in VVER-1000 Due to Large-Break Loss-of-Coolant Accident and Station Blackout Transient Using RELAP5/SCDAPSIM/MOD3.5

It is crucial to do safety evaluation of different postulated transient scenarios in actual nuclear power plants (NPPs). Some of the common analyzed scenarios are primary coolant tube rupture and station blackout (SBO). In this research, it was supposed that after establishment of a steady-state condition, an instantaneous guillotine large-break loss-of-coolant accident (LB-LOCA) of 850-mm inside diameter in one of the reactor vessel cold legs occurred, accompanied with SBO. The event progression and the variation of different reactor parameters like loop pressures, mass flow rates, fuel and clad temperature, injection rate of accumulators (ACCs), decay, and reactor power were investigated using the RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program. The reactor consequences due to availability and unavailability of passive ACCs were compared. These kinds of analyses assist in estimating the time available to perform operator safety actions. This in turn aids in emergency planning and severe accident management. The results reveal that fuel damage decreased after the introduction of ACCs. Actuation of ACCs at their actuation setpoints provided core cooling by injecting water into the reactor core. However, ACCs alone are inadequate to contain long-term core cooling during a persistent LB-LOCA. The results obtained in the research were compared with MELCOR 2.1 and ASTEC V1.3, and a cohesive agreement was obtained. Therefore, RELAP5/SCDAPSIM/MOD3.5 is capable of modeling a LB-LOCA and SBO in VVER-1000, and it provides a significant analytical capability of safety systems for specialists in the field in NPP safety.

Numerical Analysis for Hydrogen Flame Acceleration during a Severe Accident Initiated by SBLOCA in the APR1400 Containment

We performed a hydrogen combustion analysis in the Advanced Power Reactor 1400 MWe (APR1400) containment during a severe accident initiated by a small break loss of coolant accident (SBLOCA) which occurred at a lower part of the cold leg using a multi-dimensional hydrogen analysis system (MHAS) to confirm the integrity of the APR1400 containment. The MHAS was developed by combining MAAP, GASFLOW, and COM3D to simulate hydrogen release, distribution and combustion in the containment of a nuclear power plant during the severe accidents in the containment of a nuclear power reactor. The calculated peak pressure due to the flame acceleration by the COM3D, using the GASFLOW results as an initial condition of the hydrogen distribution, was approximately 555 kPa, which is lower than the fracture pressure 1223 kPa of the APR1400 containment. To induce a higher peak pressure resulted from a strong flame acceleration in the containment, we intentionally assumed several things in developing an accident scenario of the SBLOCA. Therefore, we may judge that the integrity of the APR1400 containment can be maintained even though the hydrogen combustion occurs during the severe accident initiated by the SBLOCA.