Comparative Study on Reactor Pressure Vessel Failure Behaviors With Various Geometric Discontinuities Under Severe Accident

2017 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianwei Zhu ◽  
Jianfeng Mao ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Comparative Study ◽  
Pressure Vessel ◽  
External Surface ◽  
Safety Margin ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Wide Range ◽  
Core Meltdown ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a basic strategy for severe accident (SA) mitigation of some advanced nuclear power plants (NPPs). The IVR strategy is to keep the reactor pressure vessel (RPV) intact under SA like core meltdown condition. During the IVR, the core melt (∼1327 °C) is collected in the lower head (LH) of the RPV, while the external surface of RPV is submerged in the water. Through external cooling of the RPV, the structural integrity is assumed to be maintained within a prescribed period of time. The maximum thermal loading is referred to critical heat flux (CHF) on the inside, while the external surface is considered to perform in the environment of the boiling crisis point (∼130 °C). Due to the high temperature gradients, the failure mechanisms of the RPV is found to span a wide range of structural behaviors across the wall thickness, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Besides CHF, the pressurized core meltdown was another evident threat to the RPV integrity, as indicated in the Fukushima accident on 2011. In illustrating the effects of internal pressures and individual CHF on the failure behaviors, three typical RPVs with geometric discontinuity caused by local material melting were adopted for the comparative study. Through finite-element method (FEM), the RPV structural behaviors were investigated in terms of deformation, stress, plastic strain, creep, and damage. Finally, some important conclusions are summarized in the concluding remark. Such comparative study provides insight and better understanding for the RPV safety margin under the IVR condition.

Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure

2016 ◽  
Vol 139 (2) ◽  
Author(s):  
Jianfeng Mao ◽  
Jianwei Zhu ◽  
Shiyi Bao ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
Heat Flux ◽  
Internal Pressure ◽  
Critical Heat Flux ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Wide Range ◽  
Depth Analysis ◽  
Structural Behaviors ◽  
Reactor Pressure

The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

The Influence of Crust Layer on Reactor Pressure Vessel Failure Under Pressurized Core Meltdown Accident

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Jianfeng Mao ◽  
Shiyi Bao ◽  
Zhiming Lu ◽  
Lijia Luo ◽  
Zengliang Gao
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Melt Pool ◽  
Crust Layer ◽  
Accident Management ◽  
Core Meltdown ◽  
Reactor Pressure

The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.

scholarly journals Comparative Analysis of Severe Accident at WWER-1000 NPP with MELCOR 1.8.5 and 2.1 Code Versions

2020 ◽  
pp. 30-40
Author(s):  
O. Kotsuba ◽  
Yu. Vorobyov ◽  
O. Zhabin ◽  
D. Gumenyuk
Keyword(s):  
Comparative Analysis ◽  
Pressure Vessel ◽  
Reactor Core ◽  
Computer Code ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Bottom Part ◽  
Control Volumes ◽  
Vessel Bottom ◽  
Reactor Pressure

An overview of the main improvements in updated version 2.1 of MELCOR computer code related to more representative mathematical modeling of complex thermohydraulic severe accident processes of core degradation, transfer of molten fragments to the bottom of the reactor, heating and failure of the bottom of the reactor pressure vessel is presented. The elements of WWER-1000 NPP computer model for the MELCOR 1.8.5 (control volumes, thermal structures and structures of the reactor core) that are reproduced for a reactor with the primary side, the secondary side and the containment are described. The changes implemented in WWER-1000 NPP model for MELCOR 1.8.5 to convert it to MELCOR 2.1 version that are mainly related to more detailed modeling of the reactor core and reactor pressure vessel bottom are provided. The paper presents the results of comparative analysis of severe accident scenario of total station blackout at WWER-1000 NPP with MELCOR 1.8.5 and 2.1. The comparison demonstrates good agreement between the main parameters’ results (pressure and temperature in hydraulic elements of the primary, secondary sides and the containment, temperature of core elements, the mass of the generated non-condensed gases and their concentration in the containment) obtained with these code versions for severe accident in-vessel phase. The identified differences in the time of core structures degradation and reactor vessel bottom failure are insignificantly affected by the behavior of the parameters in the primary side and the containment in the in-vessel phase of the severe accident and are related to more detailed modelling of the reactor core and bottom part of the reactor in MELCOR 2.1.

scholarly journals A Data Scatter for a Shift of the Ductile to Brittle Transition Temperature for WWER-1000 Reactor Pressure Vessel Materials

2018 ◽  
pp. 27-30
Author(s):  
V. Revka
Keyword(s):  
Transition Temperature ◽  
Standard Deviation ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Safety Margin ◽  
Temperature Shift ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Data Scatter ◽  
Reactor Pressure

In the most countries that operate the nuclear power plants with reactor pressure vessels a safety margin accounting a data scatter is applied for a conservative evaluation of a radiation shift of the ductile to brittle transition temperature for RPV metal. This scatter is to a significant extent due to material inhomogeneity and errors in determining the temperature shift and neutron fluence. In the regulatory practice of Ukraine, the obsolete approaches are used that can lead to an underestimation or overestimation of the transition temperature shift depending on the number of test data points. In order to use the updated regulatory approaches that will be consistent with international practice, it is necessary to know the magnitude of the data scatter on the transition temperature shift which is characterized by a standard deviation. Therefore, the aim of the research work was to estimate the data scatter for WWER reactor pressure vessel materials using statistical methods. The paper presents the results of a statistical analysis for a large array of surveillance test data for WWER-1000 reactor pressure vessels of NPP units which are operated in Ukraine. The data scatter for RPV base and weld metal has been estimated using a statistical treatment for the dependencies of a transition temperature shift, ΔTF, on the fast (Е > 0,5 MeV) neutron fluence. The ΔTF values have been derived from the Charpy impact tests. The Charpy V-notch specimens have been irradiated in the nuclear power reactors within a neutron fluence range of (3,0 ÷ 92,2)·1022 m-2 in the frame of a national surveillance program. The analysis has shown the data scatter relative to the average regression line for RPV materials is characterized by a standard deviation of 5,5 °С. Based on the results obtained, it was suggested to use a double standard deviation of 11 °С as a safety margin to provide a conservative estimate for the radiation shift of the transition temperature of the WWER-1000 reactor pressure vessel materials.

Investigations on the Melt Gate Ablation by Ex-Vessel Core Melts in the KAPOOL Experiments

2002 ◽  
Author(s):  
Beatrix Eppinger ◽  
Silke Schmidt-Stiefel ◽  
Walter Tromm
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Light Water Reactors ◽  
Light Water ◽  
Theoretical Tool ◽  
Transient Experiments ◽  
Core Meltdown ◽  
Water Reactors ◽  
Radioactive Release ◽  
Reactor Pressure

In future Light Water Reactors (LWR) containment failure should be prevented even for very unlikely core meltdown sequences with reactor pressure vessel (RPV) failure. In the case of such a postulated core meltdown accident in a future LWR the ex-vessel melt shall be retained and cooled in a special compartment inside the containment to exclude significant radioactive release to the environment. In such a case, a gate has to be designed to allow the melt release from the reactor cavity into the compartment. A series of transient experiments has been performed to investigate the melt gate ablation using iron and alumina melts as a simulant for the corium melt. The results of the KAPOOL tests are analyzed with the HEATING5 code in order to evaluate realistic cases of internally heated corium melts and melt gates with the same theoretical tool.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

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