Analysis of Liquid Sloshing in an In-Containment Refueling Water Storage Tank for AP1000

2013 ◽  
Author(s):  
Yu Liu ◽  
Daogang Lu ◽  
Junjie Dang ◽  
Licun Wu ◽  
Wenhui Ma
Keyword(s):  
Storage Tank ◽  
Nuclear Power ◽  
Cooling System ◽  
Water Storage ◽  
Sine Wave ◽  
Liquid Sloshing ◽  
Normal Operation ◽  
Water Storage Tank ◽  
Core Cooling ◽  
Residual Heat

Although much work has been performed on the liquid sloshing inside simple structures like rectangular and cylindrical vessels, this paper deals with the analysis of the liquid sloshing in a more complex structures, the in-containment refueling water storage tank (IRWST). The IRWST is an important component of AP1000 passive core cooling system to ensure the safe operation of the AP1000 nuclear power plant. In postulated non-LOCA events, the water in the IRWST absorbs the residual heat then transfers the heat into the containment atmosphere. However, in the case of earthquake, the sloshing fluid may influence the safety of the appropriate semi-cylindrical IRWST. In this paper, the liquid transient response in the IRWST was formulated based on finite element modal analysis when the three resonance sine wave was applied as excitation. The result shows that the maximum wave excited by excitation from different directions always emerges from the corner or the edge of tank. Another finding is that water will not overflow at the normal operational water level when exited by the selected excitations in any directions. The safety of the IRWST’s roof is achieved which guarantees that the water vapor and radioactive gases within the tank during normal operation will not release to atmosphere in the containment. The influence of the excitation direction and the water depth are also analyzed.

Seismic Response of a Cylindrical Water Storage Tank of Nuclear Power Plant

2018 ◽  
Author(s):  
Shinichi Matsuura ◽  
Ichiro Tamura
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Storage Tank ◽  
Nuclear Power ◽  
Water Storage ◽  
Response Behavior ◽  
Water Storage Tank ◽  
Elastic Plastic ◽  
Seismic Motion ◽  
Side Shell

It is important in the confirmation of the safety of the nuclear power plant to clarify the response behavior of a vertical cylindrical water storage tank under seismic motion. When a vertical cylindrical tank is shaken by a large earthquake, deformation of side shell due to the elephant foot buckling, the oval vibration etc. may occur. The occurrence of those deformations depends on materials, shapes, stored water level and time history of seismic motion. Then, response behavior was obtained for a condensate storage tank (CST) model under large seismic motion such as standard earthquake Ss multiplied by 2 with the elastic-plastic finite element calculation. In this calculation, dynamic water pressure and elastic-plastic characteristics of the material were taken into account. In this case, the elephant foot bulge did not occur but the oval vibration of side shell became dominant. Based on the result, we estimated the structural integrity of the tank.

scholarly journals AP1000 Shield Building Dynamic Response for Different Water Levels of PCCWST Subjected to Seismic Loading considering FSI

2015 ◽  
Vol 2015 ◽  
pp. 1-8 ◽  
Author(s):  
Daogang Lu ◽  
Yu Liu ◽  
Xiaojia Zeng
Keyword(s):  
Storage Tank ◽  
Seismic Analysis ◽  
Cooling System ◽  
Water Storage ◽  
Seismic Loading ◽  
Water Levels ◽  
Von Mises Stress ◽  
Fe Model ◽  
Water Storage Tank ◽  
Von Mises

Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI) under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST) placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future.

Thermal modelling of a night sky radiation cooling system

2005 ◽  
Vol 16 (2) ◽  
pp. 20-31 ◽  
Author(s):  
RT Dobson
Keyword(s):  
Storage Tank ◽  
Cooling System ◽  
Water Storage ◽  
Design Tool ◽  
Thermal Modelling ◽  
Water Storage Tank ◽  
System Components ◽  
Mathematical Equations ◽  
Night Sky ◽  
Radiation Cooling

The thermal modelling of a night sky radiation cooling system suitable for a room situated in the Namib Desert at Gobabeb, Namibia, is considered in this paper. The system consists of the following components: radiator panels, a single water storage tank, room air-to-water natural convection heat exchangers or convectors, circulating pump(s), interconnecting pipe work and temperature sensors and controls. The mathematical equations describing the thermal behaviour of the various system components are given. These equations are solved using an Excel spreadsheet and the hourly panel surface, water storage tank and room temperatures are calculated for a given internally generated heat load and weather pattern. Given the maximum allowable room temperature, the sizes of the system components may be calculated. The results obtained compared favourably with values reported in the literature. It is thus concluded that the thermal model presented can be used with confidence as a design tool for the sizing of a night sky radiation cooling system.

Analysis of Flowing Characteristics in IRWST Under Pressure Relief Condition

2013 ◽  
Author(s):  
Mengchao Zhang ◽  
Xiangbin Li ◽  
Xiaolu Fang ◽  
Yuhao Zhang
Keyword(s):  
Heat Transfer ◽  
High Pressure ◽  
Storage Tank ◽  
Nuclear Power ◽  
Water Storage ◽  
Heat Transfer Mechanism ◽  
Heat Transfer Characteristics ◽  
Pressure Relief ◽  
Water Storage Tank ◽  
Thermal Changes

The presurizer is one of main equipments of PWR Nuclear Power Plant. The heat transfer mechanism will change with temperature’s increasing when the steam with high temperature and high pressure condensates as it sprays into the in-containment refueling water storage tank (IRWST), which will be detrimental to the safety of the reactor. In this study, the flow field and heat transfer characteristics are simulated by means of professional CFD software with k-ε turbulence model and Particle modle when the steam sprays into IRWST through the sprayer. The results show thermal changes of water and steam.

Carbon Fiber Reinforcement of a Water Storage Tank for Beyond Design Basis Loads

2016 ◽  
Author(s):  
Alton Reich ◽  
John Charest
Keyword(s):  
Carbon Fiber ◽  
Storage Tank ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Storage ◽  
Safety Margin ◽  
Repair System ◽  
Buckling Mode ◽  
Water Storage Tank ◽  
Design Basis

The severe damage to the Fukushima nuclear plant occurred as a result of a beyond design basis event. This has prompted a systematic review of safety critical systems at US nuclear power plants to evaluate the existing safety margin based on beyond design basis loads. At one US nuclear power plant it was found that the Refueling Water Storage Tank (RWST) did not have sufficient margin to withstand the defined beyond design basis seismic event. An analysis indicated that the RWST would fail in an elephant foot buckling mode. This paper describes the design and analysis of a Carbon Fiber Reinforced Polymer (CFRP) repair system used to strengthen the RWST to increase the critical buckling stress for the elephant foot buckling mode.

Design of In-Containment Refueling Water Storage Tank of Chinese 3rd Generation Nuclear Power Plant

2017 ◽  
Author(s):  
Liu Yulin ◽  
Sun Xiaoying
Keyword(s):  
Storage Tank ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Storage ◽  
Stability Evaluation ◽  
Water Storage Tank ◽  
Structure Interaction ◽  
Structural Calculation ◽  
Structure Strength ◽  
The Stability

In this paper, the structure configurations of the in-containment refueling water storage tank (IRWST) of Chinese 3rd generation nuclear power plants (NPPs) was described firstly. Then, the general structural calculation for several loads, especially thermal load, were presented, as well as the stability evaluation of IRWST base-slab. The effect from fluid-structure interaction was also considered in the calculation to evaluate the design margin of IRWST. Finally, structure strength evaluation was performed for construction load case.

scholarly journals INVESTIGASI TRANSIEN TEKANAN DAN TEMPERATUR SUNGKUP REAKTOR AP1000 DALAM KECELAKAAN SBO DENGAN SET-POINT TEKANAN PENGGUYURAN BERBEDA

2015 ◽  
Vol 17 (1) ◽  
pp. 1
Author(s):  
Hendro Tjahjono
Keyword(s):  
Storage Tank ◽  
Water Storage ◽  
Heat Removal ◽  
Maximum Pressure ◽  
Set Point ◽  
Water Storage Tank ◽  
External Cooling ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Reaktor AP1000 menerapkan konsep pendinginan eksternal untuk mengantisipasi naiknya tekanan akibat terjadinya kecelakaan kehilangan seluruh catu daya listrik atau Station Black Out (SBO). Mekanisme pembuangan kalor peluruhan secara pasif dilakukan melalui Passive Residual Heat Removal System (PRHRS) yang diteruskan ke In-containment Refueling Water Storage Tank (IRWST) dan selanjutnya pada sungkup reaktor. Sungkup didinginkan secara eksternal melalui konveksi alamiah pada celah udara dan melalui penguapan air pendingin yang diguyurkan di permukaan luar dinding sungkup ketika tekanan sungkup mencapai 1,7 bar sesuai set-point yang diterapkan. Dengan mekanisme ini, tekanan akan naik sampai mencapai nilai maksimum tertentu dan kemudian turun kembali ketika pendinginan sungkup sudah mulai efektif. Tujuan dari penelitian ini adalah untuk mengetahui sejauh mana pengaruh perbedaan set-point tekanan pengguyuran tersebut terhadap tekanan dan temperatur maksimum yang dicapai. Metode yang digunakan adalah dengan melakukan simulasi menggunakan model perhitungan analitik berbasis Matlab-07 pada kondisi transien yang mampu mengestimasi daya kalor yang dievakuasi, tekanan dan temperatur di dalam sungkup terhitung mulai terbentuknya uap di dalam sungkup. Hasil simulasi menunjukkan pola transien tekanan dan temperatur yang naik hingga maksimum dan turun kembali ke suatu nilai yang relatif tetap. Dengan variasi set-point mulai dari 1,7 bar hingga 5 bar, tekanan maksimum yang dicapai meningkat dari 3,5 bar hingga 5 bar dan temperatur maksimum dari 117 °C hingga 125 °C. Dapat disimpulkan bahwa di AP 1000, dengan naiknya set-point tekanan dimulainya pendinginan eksternal melalui pengguyuran air berpengaruh menaikkan tekanan dan temperatur maksimum yang terjadi akibat SBO. Kata kunci: Transien tekanan, set-point pendinginan eksternal sungkup, AP1000, SBO.  AP1000 reactor applying external cooling concept to anticipate the increase in pressure due to Station Black Out (SBO). Disposal mechanism of decay heat conducted through the Passive Residual Heat Removal System (PRHRS) to In-containment Refueling Water Storage Tank (IRWST) and subsequently forwarded to the reactor containment. Containment is externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall when the pressure attaints 1.7 bars according to the applied pressure set-point. With this mechanism, the pressure will increase until it reaches certain maximum value and then decrease when containment cooling already begun effectively. The purpose of this study was to determine the effect of the set-point to the maximum pressure and temperature reached. The utilized method is to perform simulations using Matlab-07 model of analytical calculations based on a transient state that is capable of estimating the power of heat evacuated and the pressure in the containment. The simulation results show the pattern of pressure and temperature transient rises to a maximum and drops back to a value that is relatively constant. With the set-point variation ranging from 1.7 bars to 5 bars, the maximum pressure varies from 3.5 bars to 5 bars and the maximum temperature varies from 117 °C to 125 °C. It can be concluded that with increasing the set-point pressure of starting the external cooling with water, the maximum pressure and temperature increase. Keywords: Transient pressure, containment external cooling set-point, AP1000, SBO.

Development of Best-Estimate ECCS Evaluation Methodology for APR1400

2009 ◽  
Author(s):  
Seok-Ho Lee ◽  
Mun-Soo Kim ◽  
Han-Gon Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Evaluation Methodology ◽  
Preliminary Calculation ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
Best Estimate ◽  
Design Characteristics ◽  
Core Cooling

Advanced Power Reactor 1400 (APR1400) is an evolutionary Pressurized Water Reactor (PWR) equipped with such advanced features as the Direct Vessel Injection (DVI), the Fluidic Device (FD) in the Safety Injection Tank (SIT), and the In-containment Refueling Water Storage Tank (IRWST) in the Emergency Core Cooling System (ECCS). To verify the performance of these advanced features, more realistic performance evaluation methodology is desired since existing methodologies use too conservative assumptions which cause negative biases to these features. In this study, therefore, a best estimate evaluation methodology for the APR1400 ECCS under large break loss of cooling accident (LBLOCA) is developed targeting operating license of the Shin Kori 3&4 nuclear power plants (SKN 3&4), the first commercial APR1400 plants. On this purpose, a variety of existing best estimate evaluation methodologies previously used are reviewed. As a result of this review, a methodology named KREM is selected for this study. The KREM is based on RELAP5/MOD3.1K and has been used for Korean operating plants since 2002 when it was first approved by Korean regulation. For this study, RELAP5/MOD3.3 (Patch 3), the latest version of RELAP series is selected since it could appropriately simulate the multi-dimensional phenomena for the APR1400 design characteristics. To quantify the code accuracy, analyses covering experimental data have been performed for 36 kinds of separated effect tests (SETs) and integral effect tests (IETs). The uncertainty in the peak cladding temperature (PCT) of the APR1400 is evaluated preliminarily. Based on the preliminary calculation, final uncertainty quantification and bias evaluation are performed to obtain the licensing PCT for Shin-Kori 3&4 plants and the result shows that the LBLOCA licensing acceptance criteria are well satisfied.

Design of a Tornado Missile Resistant Shield for the Angra Nuclear Plant

2015 ◽  
Author(s):  
Alton Reich ◽  
Jason Moon
Keyword(s):  
Storage Tank ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Storage ◽  
Environmental Hazards ◽  
Water Tank ◽  
Raw Water ◽  
Water Storage Tank ◽  
Design Basis ◽  
Under Construction

The Angra site contains the only two operational nuclear power plants in Brazil and one that is under construction. When Angra Unit 1 was designed the risk of tornadoes was deemed to be vanishingly small and tornadoes were not included in the design basis for the plant. When the question of environmental hazards was revisited for the design basis of Angra Unit 3, tornadoes were included as part of the design basis. Based on this change in philosophy for the Angra site, a review was performed and it was determined that the Raw Water Storage Tank and Reactor Make-up Water Tank, both of which are located outdoors at Angra Unit 1 required protection from tornado driven missiles. The tanks contain water that may be used for emergency cooling, and the water must be available for use during a design basis event. This paper provides an overview of the methodology used to design the tornado missile resistant shield structure for these two tanks.

Experimental Study on the Performance of IIST Passive Core Cooling System

2002 ◽  
Author(s):  
Chin-Jang Chang ◽  
Chien-Hsiung Lee ◽  
Wen-Tan Hong ◽  
Lance L. C. Wang
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
General System ◽  
System Response ◽  
Water Storage Tank ◽  
The Core ◽  
Third Stage ◽  
Core Cooling

The purpose of this study is to conduct the experiments at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility for evaluation of the performance of the passive core cooling system (PCCS) during the cold-leg small break loss-of-coolant accidents (SBLOCAs). Five experiments were performed with (1) three different break sizes, 2%, 0.5%, and 0.2% (approximately corresponding to 1 1/4”, 2”, and 4” breaks for Maanshan nuclear power plant), and (2) 0.2% and 0.5% without actuation of the first-stage and third-stage automatic depressurization valve (ADS-1 and ADS-3) to initiate PCCS for assessing its capacity in accident management. The detailed descriptions of general system response and the interactions of core makeup tanks (CMTs), accumulators (ACCs), automatic depressurization system (ADS), passive residual heat Removal (PRHR), and in-containment refueling water storage tank (IRWST) on the core heat removal are included. The results show: (1) core long term cooling can be maintained for all cases following the PCCS procedures, (2) the core can be covered for the cases of the 0.2% and 0.5% breaks without actuation of ADS-1 and ADS-3.

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