The Study of CAP1400 Core Melt and Relocation for IVR Analysis

2017 ◽  
Author(s):  
Jiayun Wang ◽  
Wei Lu ◽  
Pei Wen Gu
Keyword(s):  
Complex Structure ◽  
Pool Analysis ◽  
Large Mass ◽  
Severe Accident ◽  
Retention Strategy ◽  
The Core ◽  
Melt Pool ◽  
Element Simulation ◽  
Simulation Based ◽  
Large Heat

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.

Simulation of Molten Corium Concrete Interaction With the MOCO Code

2014 ◽  
Author(s):  
Bo Lin ◽  
Sui-zheng Qiu ◽  
Guang-hui Su ◽  
Wen-xi Tian ◽  
Ya-pei Zhang
Keyword(s):  
Numerical Study ◽  
Computer Code ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Core ◽  
Melt Pool ◽  
Transfer Behavior ◽  
Cooling Behavior ◽  
Reactor Pressure

In the event of a severe accident in a pressurized water reactor, the core of a reactor melts and forms corium, a mixture that includes molten UO2 and ZrO2. If the reactor pressure vessel fails, corium can be relocated in the containment cavity and interact with concrete forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question is what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using a new computer code MOCO. The code considers the heat transfer behavior in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, and incorporates phenomenology that is deemed to be important for analyzing debris cooling behavior. The interaction between thermalhydraulics and physic-chemistry is modeled in MOCO. The main purpose of this paper is to present the modeling used in MOCO and some validation calculations using the data of experiments available in the literature.

scholarly journals Finite Element Simulation Based Analysis of Valve-sparing Aortic Root Surgery

2020 ◽  
Vol 53 (2) ◽  
pp. 16037-16042
Author(s):  
Róbert Nagy ◽  
Tamás Umenhoffer ◽  
Péter Somogyi ◽  
Ákos Szlávecz ◽  
Anikó Kubovje ◽  
...  
Keyword(s):  
Finite Element ◽  
Finite Element Simulation ◽  
Aortic Root ◽  
Element Simulation ◽  
Aortic Root Surgery ◽  
Simulation Based ◽  
Valve Sparing

A model for calculating composition and density of the core melt in the water-moderated water-cooled reactor in case of severe accident

2010 ◽  
Vol 57 (13) ◽  
pp. 1099-1111 ◽  
Author(s):  
V. D. Ozrin ◽  
O. V. Tarasov ◽  
V. F. Strizhov ◽  
A. S. Filippov
Keyword(s):  
Severe Accident ◽  
The Core

Experiment and Simulation Analysis of the Pressure Carrying Capacity of X80 Pipe with Metal Loss Defect on the Girth Weld

2021 ◽  
Vol 1035 ◽  
pp. 813-818
Author(s):  
Zheng Long Li ◽  
Lin Chen ◽  
Zhi Hong Li ◽  
Guo Shuai Yan ◽  
Wei Li
Keyword(s):  
Sensitivity Analysis ◽  
Carrying Capacity ◽  
Simulation Analysis ◽  
Water Pressure ◽  
Metal Loss ◽  
Factors Affecting ◽  
Element Simulation ◽  
Simulation Based ◽  
Blasting Test ◽  
Water Pressure Blasting

In order to study the pressure carrying capacity of X80 pipe with metal loss defect on the girth weld the water-pressure blasting test of the pipe with metal loss defect was analyzed by experiment and finite element simulation. Based on this, the sensitivity analysis of the factors affecting the pressure carrying of the pipeline, such as the circular size, the axial size, and the depth of the metal loss defect, was carried out. The research results show that the circular size of the metal loss defect on the girth weld had little impact to the pressure carrying capacity of the pipe while it reduced with the increasing of the axial size and the depth of the metal loss defect.

A Model of the Fission Products Release from the Melt Pool during Severe Accident in a Liquid Metal Cooled Fast Reactor

2021 ◽  
Vol 68 (2) ◽  
pp. 152-158
Author(s):  
E. V. Usov ◽  
V. I. Chukhno ◽  
I. A. Klimonov ◽  
V. D. Ozrin ◽  
N. A. Mosunova ◽  
...  
Keyword(s):  
Liquid Metal ◽  
Fast Reactor ◽  
Fission Products ◽  
Severe Accident ◽  
Melt Pool

Ablation of a Solid Material by High Temperature Liquid Jet Impingement: An Application to Corium Jet Impingement on a Sfr Core-Catcher

2021 ◽  
Author(s):  
Alexandre Lecoanet ◽  
Michel Gradeck ◽  
Xiaoyang Gaus-Liu ◽  
Thomas Cron ◽  
Beatrix Fluhrer ◽  
...  
Keyword(s):  
High Temperature ◽  
Deep Understanding ◽  
Jet Impingement ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Liquid Jet ◽  
Cavity Formation ◽  
The Core ◽  
Core Catcher ◽  
Temperature Liquid

Abstract This paper deals with ablation of a solid by a high temperature liquid jet. This phenomenon is a key issue to maintain the vessel integrity during the course of a nuclear reactor severe accident with melting of the core. Depending on the course of such an accident, high temperature corium jets might impinge and ablate the vessel material leading to its potential failure. Since Fukushima Daiichi accident, new mitigation measures are under study. As a designed safety feature of a future European SFR, bearing the purpose of quickly draining of the corium out of the core and protecting the reactor vessel against the attack of molten melt, the in-core corium is relocated via discharge tubes to an in-vessel core-catcher has been planned. The core-catcher design to withstand corium jet impingement demands the knowledge of very complex phenomena such as the dynamics of cavity formation and associated heat transfers. Even studied in the past, no complete data are available concerning the variation of jet parameters and solid structure materials. For a deep understanding of this phenomenon, new tests have been performed using both simulant and prototypical jet and core catcher materials. Part of these tests have been done at University of Lorraine using hot liquid water impinging on transparent ice block allowing for the visualizations of the cavity formation. Other tests have been performed in Karlsruhe Institute of Technology using liquid steel impinging on steel block.

A high-sensitivity dual-sample synchronous detection photonic crystal fiber sensor

2021 ◽  
pp. 2150306
Author(s):  
Pibin Bing ◽  
Guifang Wu ◽  
Zhongyang Li ◽  
Sheng Yuan ◽  
Hongtao Zhang ◽  
...  
Keyword(s):  
Photonic Crystal ◽  
Photonic Crystal Fiber ◽  
Complex Structure ◽  
High Sensitivity ◽  
Structure Design ◽  
Synchronous Detection ◽  
Crystal Fiber ◽  
The Core ◽  
Wavelength Sensitivity ◽  
Air Hole

The photonic crystal fiber (PCF) sensor based on surface plasmon resonance (SPR) technology has flexibility in birefringence, negative dispersion, effective area and nonlinearity and has become a widely studied new fiber. However, there are many problems in the practical application of the sensor, such as complex structure design, not easy to prepare, the low sensitivity of sensing detection, narrow range of refractive index detection of analytes, which greatly limits the application range and functions of PCF sensors. To solve the above problems, this paper adopts a structure with a longer vertical distance between the D-shaped large air hole channel and the core. The energy of the core conduction mode is better limited by the cladding to transmit in the core, and the plasma mode is enhanced, which can effectively increase the wavelength sensitivity. In this paper, the hexagonal double-clad air hole structure and the D-type structure are combined to ensure a simple structure and facilitate manufacturing and production, while the wavelength sensitivity is also greatly improved. The wavelength sensitivity of the dual sample channel can reach up to 16200 nm/RIU and 15800 nm/RIU, which has broad application prospects in the field of high-sensitivity detection.

Research on Finite Element Method of Hot Forming of High Strength Steel Sheet

2021 ◽  
Vol 1032 ◽  
pp. 172-177
Author(s):  
Xiao Da Li ◽  
Xiang Hui Zhan
Keyword(s):  
Finite Element ◽  
Finite Element Simulation ◽  
High Strength Steel ◽  
Strong Support ◽  
High Strength ◽  
Hot Forming ◽  
Mold Design ◽  
Deformation Method ◽  
Element Simulation ◽  
Simulation Based

The finite element simulation technology can provide strong support for the optimization of processing technology and the treatment of detailed problems in the processing process. Two finite element methods applied to hot forming of high-strength steel plates are introduced, namely the incremental method and the deformation method. Two methods are used for simulation calculations. The finite element simulation based on incremental theory has high accuracy and requires more complete mold and process information. It is mainly used in the middle and late stages of product and mold design. And the finite element simulation based on deformation theory have fast calculation speeds and are mainly used in the early stages of product and mold design. Both types of methods have high practical value.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

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