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.

scholarly journals Evaluation of VVER-1200/V-491 Reactor Pressure Vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

2015 ◽  
Vol 5 (4) ◽  
pp. 54-63
Author(s):  
Thi Hoa Bui ◽  
Chi Thanh Tran
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Worst Case ◽  
Safety Requirements ◽  
Vessel Integrity ◽  
Reactor Pressure

After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term.

Study of PWR hot leg creep rupture and RCS depressurization strategy during an SBO accident

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 194-201
Author(s):  
L. Wu ◽  
H. Miao ◽  
P. Yu ◽  
Z. Huang ◽  
J. Zheng ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Failure Time ◽  
Fission Products ◽  
Creep Rupture ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Relief Valve ◽  
Pressurized Water ◽  
Station Blackout ◽  
Reactor Pressure

Abstract Preventing the leakage of radioactive materials is important to nuclear safety. During a station blackout accident in pressurized water reactors, the hot leg creep rupture caused by hot leg countercurrent flow occurs before the reactor pressure vessel failure that caused by lower head rupture. The secondary fission products barrier is lost after hot leg creep rupture. An analysis for this phenomenon was done using the Modular Accident Analysis Program version 4.0.4 code. A station blackout accident for CPR1000 is simulated and the occurrence and influence of hot leg creep rupture phenomenon are analyzed in detail. After that, a sensitivity analysis of the opening of different pressurizer pilot-operated relief valves at five minutes after entering severe accident management guideline (before the hot leg creep rupture occurs) is studied. The results show that reactor pressure vessel failure time can be extended by at least 4 h if at least one pilot-operated relief valve is opened and direct containment heating phenomenon can be eliminated if at least two pilot-operated relief valves are opened.

scholarly journals Influence of Modelling Options in RELAP5/SCDAPSIM and MAAP4 Computer Codes on Core Melt Progression and Reactor Pressure Vessel Integrity

2010 ◽  
Vol 2010 ◽  
pp. 1-11 ◽  
Author(s):  
Siniša Šadek ◽  
Srđan Špalj ◽  
Bruno Glaser
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Primary System ◽  
Coolant Pump ◽  
The Core ◽  
Hydraulic Behaviour ◽  
Computer Codes ◽  
Vessel Integrity ◽  
Reactor Pressure

RELAP5/SCDAPSIM and MAAP4 are two widely used severe accident computer codes for the integral analysis of the core and the reactor pressure vessel behaviour following the core degradation. The objective of the paper is the comparison of code results obtained by application of different modelling options and the evaluation of influence of thermal hydraulic behaviour of the plant on core damage progression. The analysed transient was postulated station blackout in NPP Krško with a leakage from reactor coolant pump seals. Two groups of calculations were performed where each group had a different break area and, thus, a different leakage rate. Analyses have shown that MAAP4 results were more sensitive to varying thermal hydraulic conditions in the primary system. User-defined parameters had to be carefully selected when the MAAP4 model was developed, in contrast to the RELAP5/SCDAPSIM model where those parameters did not have any significant impact on final results.

Analysis on Manufacturing Quality Control of 1000 MWe Reactor Pressure Vessel

2013 ◽  
Vol 803 ◽  
pp. 442-447
Author(s):  
Hong Wei Cai ◽  
De Peng Guo ◽  
Xiao Ming Qian ◽  
Peng Zhang
Keyword(s):  
Quality Control ◽  
Statistical Analysis ◽  
Pressure Vessel ◽  
Manufacturing Process ◽  
Nuclear Power ◽  
Basic Problem ◽  
Reactor Pressure Vessel ◽  
Manufacturing Quality ◽  
The Core ◽  
Reactor Pressure

Reactor Pressure Vessel is the core equipment in the nuclear power plant, many problems happen during the manufacturing process of localization. Through the statistical analysis on non-conformance and problems detected by quality surveillance, the characteristics and difficulties of RPV quality control in domestic manufacturing process is researched, and the basic problem and the direction of improvement has been analyzed. The data, contents and opinions will further provide reference for RPV and other similar equipments manufacturing quality control.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

2021 ◽  
Vol 14 (1) ◽  
pp. 34-39
Author(s):  
D. A. Kuzmin ◽  
A. Yu. Kuz’michevskiy
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Practical Significance ◽  
Reactor Unit ◽  
Operating Modes ◽  
Brittle Destruction ◽  
Reactor Pressure

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

Reactor Pressure Vessel Modification Approaches

2005 ◽  
Author(s):  
Ronald J. Payne ◽  
Stephen Levesque
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
State Of The Art ◽  
Electrical Power ◽  
Evaluation Criteria ◽  
Reactor Pressure Vessel ◽  
Alloy 600 ◽  
Electrical Power Research Institute ◽  
Reactor Pressure ◽  
Plant Components

Stress corrosion cracking of Alloy 600 has lead to the modification and replacement of many nuclear power plant components. Among these components are the Bottom Mounted Nozzles (BMN) of the Reactor Pressure Vessel (RPV). Modifications of these components have been performed on an emergent basis. Since that time, Framatome ANP has developed state-of-the-art modification methods for the repair of BMNs using the Electrical Power Research Institute (EPRI) managed Materials Reliability Program (MRP) attributes for an ideal repair as a basis for evaluation of modification concepts. These attributes were used to evaluate the optimal modification concepts and develop processes and tooling to support future modification activity. This paper details the BMN configurations, modification evaluation criteria, several modification concepts, and the development of the tooling to support the optimal modification scenarios.

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