LOCA Accident Research Articles

Thermal safety characteristics analysis of helium xenon gas cooled small reactor system under LOCA accidents

The thermal safety characteristics of the system after a small break loss of coolant accident (SB-LOCA) in a helium xenon gas-cooled small reactor will determine whether the reactor will experience a meltdown and the duration of the meltdown. At present, research on SB-LOCA both domestically and internationally is mostly focused on water-cooled reactors, lacking relevant research on helium xenon gas-cooled reactors (He-Xe GCR). Therefore, it is of great significance to research the thermal safety characteristics of He-Xe GCR after a breach loss of coolant accident. This article improves the gas turbine model based on the original SB-LOCA analysis program for pressurized water reactors. And added a calculation model for the characteristics of helium xenon mixed gas, a heat transfer model for helium xenon mixed airflow, and a compressor model. Finally, a Brayton Cycle Reactor System Analysis program (BRESA) was developed for SB-LOCA analysis of He-Xe GCR. Explore the impact of the size and location of fractures on the thermal safety characteristics of the system using BRESA. Research has found that as the size of the rupture gradually increases, the decrease in reactor core power is greater, and the reactor power is shut down earlier. Comparative analysis of the impact of different breach positions on the residual heat removal characteristics reveals that a breach accident on the high-pressure side of the system poses a greater threat to system safety. Therefore, when setting the safety threshold for the Brayton cycle system, the focus should be on the hazards of high-voltage side breaches. This research achievement assists with the safe operation strategy of He-Xe GCR in the future.

Identification of Cold-Leg Break Size in LOCA Accident using Artificial Neural Networks and Simulation Database

The most widely studied LOCA (Loss of Coolant Accidents) is a rupture of a cold leg pipe causing the Reactor Cooling System to depressurize first, with different break sizes corresponding to the change in trigger signal from the Instrument and Control System (I&C System) such as pressure, temperature, power, pressure vessel water level, etc. is different. Therefore, the response of nuclear power plant varies considerably with the size of break. To mitigate the consequence of LOCA with a given break size, it is necessary to design the emergency core coolant systems so that the fuel is cooled efficiently during all phases of the accident. Therefore, the size of rupture needs to be detected and identified as soon as possible right after reactor scram. To achieve this goal, this study is conducted to investigate the applicability of artificial neural networks (ANN) for recognizing LOCAs, especially identifying the rupture sizes of the LOCAs according to the changes of operational parameters of VVER-1000 nuclear power plant. This study mainly focuses on building, training, and optimizing the artificial neural networks using simulation databases obtained from the RELAP5 simulation program for VVER-1000 reactor technology. Results clearly showed the potential application of ANN-based model for detecting the break size even with uncertainty of input parameters added.

CFD investigation of PWR channel shapes effects during rolling motion under normal and accident condition

Thermal-hydraulic parameters of the floating reactor fluctuate constantly, which can endanger the safety features and make challenges for safety systems. The rolling motion has more effect on the thermal-hydraulic of the floating reactor between other six motions. In this study, the effect of fuel shape and heated channel shape on thermal-hydraulic parameters were evaluated along with analysis of the parameter value changes during rolling motion under normal and LOCA accident. Also, simulation of the rolling motion with the oscillating velocity is done. The cylindrical and rectangular fuel shapes, as well as hexagonal and square channel shapes, were compared. For modeling the rolling motion, an UDF program is developed for solving the related equations in CFD code. The results indicated that the rectangular fuel with the square channel has lower changes during rolling motion than the other two models, which can be considered as the main reason for choosing the plate or rectangular fuel types in the new floating reactors.

Experiment Investigation of a Secondary Prhr Cooling Effect on an Integral Facility after a Loca Accident

Experiment Investigation of a Secondary Prhr Cooling Effect on an Integral Facility after a Loca Accident

МОДЕЛИРОВАНИЕ УНОСА КАПЕЛЬ В СТРУЙНО-ВИХРЕВОМ КОНДЕНСАТОРЕ СИСТЕМЫ ЛОКАЛИЗАЦИИ АВАРИЙ ВВЭР-440

The article presents the results of simulating the jet-vortex condenser operation. The jet-vortex condenser used as a part of the confinement system is designed to ensure the confinement integrity of an NPP with VVER-440 unit during LOCA accidents. To simulate the jet-vortex condenser operation it is important to keep a sufficient amount of water after a pressure peak was reached in the jet-vortex condenser hydraulic lock. The loss of water due to drop entrainment with steam-air mixture flow into the atmosphere stops when the velocity of drop sedimentation becomes higher than the velocity of drop entrainment with the mixture flow. The jet-vortex condenser model integrated into the KUPOL-M code was validated against the experimental data obtained on the VNIIAES test facility. To take into account drop entrainment with steam-air mixture flow the procedure of moisture separation and drop entrainment was used. A good agreement between calculated and experimental results was obtained when comparing the initial and final water levels in the hydraulic lock. The research results confirmed validity of the model of drop entrainment with steam-air mixture flow during the operation of the jet-vortex condenser and the preservation of water into the hydraulic lock during the accident.

Conceptual design of reactor system for hybrid micro modular reactor (H-MMR) using potassium heat pipe

Hybrid micro modular reactor (H-MMR) combining MMR with renewable energy and energy storage systems (ESS) is designed using heat pipe. H-MMR could produce 10 MW of electric power more flexibly and efficiently through load following. LOCA accident is precluded in advance as there is no primary coolant due to the use of potassium heat pipe. Reactor solid core consist of hexa-annulus UN fuel with a heat pipe inserted, which connects a sodium pool of intermediate heat exchanger (IHX). The annular gap wick heat pipe (ACHP) using potassium is optimized to maximize the heat transport performance. Reactor core configuration is designed through core analysis and thermal analysis using 1D lumped finite differential method. Reactor core of H-MMR make 18 MW of thermal power for 56 years without refueling when diameter of heat pipe is 22 mm. The feasibility of heat pipe safety system and sodium pool is evaluated in normal operation. Reactor vessel auxiliary cooling system (RVACS) is designed to cool down decay heat within safety margin using implicit transient analysis.

The measure on mitigating hydrogen risk during LOCA accident in nuclear power plant

Since there is a large empty space in PWR containment which has strong pressure bearing capability, it’s generally considered that there is no overall risk of hydrogen. However, the complex internal structure makes hydrogen easy to accumulate, burn or even explode in local compartment. Therefore, when LB-LOCA with gravity injection failure happens, there may be a more serious risk of local hydrogen.GASFLOW, a three-dimensional computational fluid dynamics code, is used to study the measure on mitigating hydrogen risk. During the simulation, a steam injection device is arranged in the PZR compartment. When the accident occurs, steam is injected into the PZR compartment during the pre-set stages. By comparing the hydrogen risks in different situations, it could be found that steam injection in local space can reduce the hydrogen risk locally. The method combining steam injection device with PARs is an effective measure to mitigate the risk of hydrogen.

Gamma and neutron dosimetry of Tehran Research Reactor containment during and after LOCA accident

The radiation effects of loss of coolant accident on the reactor operators are critical issues. In the present study, computational codes were used to investigate gamma and neutron dose rates inside the 5 MW pool-type Tehran Research Reactor containment. The carried out calculations showed, when the remaining coolant over the core is decreasing less than 1 m, the containment should be evacuated. In addition, 24 h after LOCA, the containment gamma dose rates drop to one tenth of their initial values immediately after LOCA.

Simulations for cooling effect of PCCS in hot leg SB-LOCA of 1000 MW PWR

The passive containment cooling system (PCCS) for reactor containment is a security system that can be used to cool the atmosphere and reduce pressure inside the containment in case of temperature and pressure increase caused by vapor injection, which requires no external power since it works only with natural forces. This article aims to establish a CFD simulation model for the Passive Containment Cooling System of 1000MW PWR using Code_Saturne and FLUENT software. The comparison of 4 different models based respectively on mixture model, COPAIN test, Uchida correlation and Chilton-Colburn analogy are used to simulate the condensing effect, the improvement of source code are based on a 3D simulation of PCCS system. To simulate the thermal-hydraulic condition in the containment after LOCA accident caused by a double-ended main pipe rupture, a high temperature vapor with the given mass flow rate are supposed to be the source of energy and mass. Meanwhile the surface of three condensing island use the wall condensation model.The simulation results obtained from the 4 models show similar transient process, the difference between the steady-state analysis of three models is less than 3%. Meanwhile, the large mass flow rate of water loss inside the containment induce a high flow rate of steam which could be uniformly mixed with air in a short time, the impact of steam injection to different directions on the simulation results is not obvious. For the self-condensing efficiency of 3 groups of PCCS system, the non-centrosymmetric injection position resulting that the condensing efficiency is slightly higher for the two heat exchanger groups nearby. During the first 2400s of simulation time, more than 75.69% of the steam is condensed, the occurrence of condensation at the wall mainly driven by natural convection, the effect of thermodynamic siphon could improve the flow rate of gas mixture inside the tubes when the velocity of mixture is not large enough. The steam could smoothly enter the tube and reach the internal cooling surface then to be condensed. Besides, PCCS ensure that the containment internal pressure is maintained below 2bar and the temperature is maintained below 380K during 3600s.

Effective water cooling of very hot surfaces during the LOCA accident

The quench front propagation in the case of the water emergency cooling of water cooled fusion nuclear components is studied during the LOCA accident. In the case of the LOCA accident, heat removal can be successful only when water rewetts the surface. The place where water meets the very hot surface is called the quench front. The quench front propagation depends on the initial wall temperature, coolant flow rate, and heat accumulated in the cooled components. Results of the quench front experiment in an annular 1.7m length channel with initial wall temperatures within the range from 250°C up to 800°C and coolant mass fluxes 100kg/m2s and 180kg/m2s are presented. From the results is concluded, that quenching velocities and rewetting temperatures are strongly dependent on the initial wall temperature as well as the pressure spikes. The behaviour of the heat transfer coefficient at the quench front is analyzed.

QUENCH FRONT PROPAGATION IN THE ANNULAR CHANNEL

Understanding the quench front propagation during bottom core reflooding is crucial for the effective cooling during the LOCA accident. The results presented in this paper were obtained on an experimental loop with an annular test section. The test section consists of a vertical electrically heated stainless steel tube with outer diameter 9 mm and length of 1.7 m. The heated tube is placed inside a glass tube with the inner diameter 14.5 mm. Water mass flux during the reflooding is in the range from 100 kg.m<sup>−2</sup>.s<sup>−1</sup> up to 140 kg.m<sup>−2</sup>.s<sup>−1</sup> and the initial wall temperature of the stainless steel tube is in the range from 250 °C up to 800 °C. The presented results show the influence of the initial conditions on the quench front propagation and the complexity of the phenomenon.

Cooling the intact loop of primary heat transport system using Shutdown Cooling System in case of LOCA events

The purpose of this paper is to model the operation of the Shutdown Cooling System (SDCS) for CANDU 6 nuclear power plants in case of LOCA accidents, using Flowmaster calculation code, by delimiting models and setting calculation assumptions, input data for hydraulic analysis and input data for calculating thermal performance check for heat exchangers that are part of this system.

Npp Safety in Slovakia According to Stress Tests After Accident in Fukushimi

Abstract This paper presents the new requirements to test of the safety and reliability of the NPP structures due to the last nuclear accidents in the world. The accidents of the NPP in Chernobyl and Fukushima give us the new inspiration to verify the safety level of the NPP structures. The probabilistic assessment of NPP structures for PSA level 2 of VVER 440 in the case of LOCA accident is presented. The results of the probabilistic nonlinear analysis of the NPP structures are presented.

Analysis on containment depressurization under severe accidents for a Chinese 1000 MWe NPP

Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment.

Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety.

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

Dynamic response of a nuclear fuel rod impacting on elastoplastic gapped supports

The paper concerns the study of the vibrational behavior of a fuel rod supported by elastoplastic supports. Nuclear fuel rods in reactor conditions are always exposed to turbulent flow, which in turn creates random vibrations. Shock vibration can also occur as a result of unwanted conditions, such as seismic and LOCA accidents. Excessive external forces can induce plastic deformation in the nuclear fuel rod. Therefore, a bilinear elastoplastic contact force model is utilized to simulate nonlinear behavior, and the plastic deformation of the supports is computed based on a predictor–corrector scheme. Once the contact force exceeds the yielding point, it can be shown that the supports behave nonlinearly, and that the gap becomes larger due to the plastic deformation. Three categories of vibration – harmonic, random, and shock vibration – are dealt with in this work. The elastoplastic response of supports subjected to such vibrations is reviewed, and the stress levels developed in the fuel cladding is evaluated. It has been found that the minimum required force to cause plastic deformation is dependent on excitation frequency as well as gap size, but it can be seen that the excitation frequency range is more sensitive for random vibrations. For all simulation cases, the stress level of the fuel cladding is below the yield limit, and results show that the most severe conditions are induced by random vibrations in terms of the stress intensity.

Probabilistic Safety Analysis of the Nuclear Power Plants in Slovakia

Probabilistic Safety Analysis of the Nuclear Power Plants in Slovakia This paper gives the results of the safety analysis of the nuclear power plants in Slovakia. The probabilistic assessment of NPP concrete containment and technologically steel segment penetration for Probabilistic Safety Analysis (PSA) level 2 of VVER 440/213 in the case of seismic and LOCA accident is presented. There is showed summary of calculation models and calculation methods for the probability analysis of the structural integrity considering load, material and model uncertainties. The numerical simulations on the base of LHS method were realized in the system ANSYS and FReET. Copyright©2010 Preprint of KONBiN

Surface reaction of titanium beryllide with water vapor

Beryllium is one of the candidate materials for the neutron multiplier in the tritium-breeding blanket. However, there are some problems related to the application of beryllium as a neutron multiplier, which include compatibility with structural materials, tritium inventory and reaction of beryllium with water vapor and oxygen in a LOCA accident. Titanium beryllides such as Be 12Ti are known to have advantages over beryllium from the perspectives of higher melting point, lower chemical reactivity and lower swelling. Thus, these materials are promising alternatives of beryllium. The authors investigate here the reaction of titanium beryllides with water vapor at high temperatures, and it is found that Be 12Ti is by far more tolerant to water vapor than beryllium. To clarify the high tolerance of Be 12Ti to water vapor, the surface of Be 12Ti used in the experiment was investigated by means of digital microscope, SEM, XRD and electron spectroscopy for chemical analysis (ESCA). These analyses suggest that some oxidized state of beryllium was formed on the surface of Be 12Ti exposed to water vapor. In particular, the results of the ESCA analysis gave some clues to understanding the higher tolerance of Be 12Ti to water vapor at high temperatures.

Dynamic response analysis for fuel assemblies under earthquake and LOCA accidents

The fuel assembly is the main component in 300 MWe PWRs, located in an environment of high temperature, high pressure, irradiation and erosion of the reactor core. Whether or not it can keep functioning during the normal operation of plants, especially under severe accidents such as the earthquake and LOCA, will affect directly the safety of the NPPs. In this paper, by using the general-use computer code ANSYS, a horizontal modal analysis model and a non-linear impact analysis model as well as a vertical non-linear impact analysis model for the fuel assemblies are constructed. The results of the modal analysis are compared with that obtained from tests and other computer codes. Relevant parameters in the non-linear impact calculating models are selected and adjusted by comparing with the shaking table test data of the fuel assemblies. Based on the comparisons and adjustments, the dynamic response calculations of the 1×13 horizontal impact models for Qinshan and CHASNUPP under the earthquake (SSE) and LOCA conditions are performed and corresponding results are given. The response calculations of the vertical impact model for three kinds of postulated LOCA cases are also performed and relevant calculated results are provided.