scholarly journals Results of Structural Analysis of WWER-1000/320 Containment Behavior under Severe Accidents

2020 ◽  
pp. 18-30
Author(s):  
L. Liashenko ◽  
A. Panchenko ◽  
O. Shugailo ◽  
M. Koliada
Keyword(s):  
Strain State ◽  
Cylindrical Part ◽  
Severe Accident ◽  
Normal Operation ◽  
Stress And Strain ◽  
Severe Accidents ◽  
Design Basis ◽  
Calculation Results ◽  
Design Basis Accident ◽  
Maximum Design

The paper presents the review and evaluation of the containment prestressing system within reinforced concrete structures under seismic loads and severe accidents. Given the complex design of the containment, the detailed finite element model has been developed and used to describe real containment behavior. Containment stress and strain state was calculated by modern LIRA software. The first stage analyzed the results of WWER-1000/320 containment stress and strain state calculation under a combination of loads caused by maximum design basis accident (MDBA) and safe shutdown earthquake (SSE) and defined minimum acceptable tension of tendons. The research determines the minimum acceptable tension of tendons in the containment prestressing system, and evaluates the strength and reliability of containment structures under a combination of loads in normal operation + design-basis accident + maximum design earthquake (NO + DBA + MDE). The verification calculations have been performed using tendon tension of 780 ton-force in the cylindrical part of the containment and 760 ton-force in the containment dome. The second stage covered the analysis of severe accident parameters (pressure and temperature) and the results of calculation. Stress and strain state in ZNPP-1 containment has been calculated, parameters (pressure and temperature) under which the containment can loss its protective and isolation functions have been identified, calculation results have been analysed and conclusions of containment structural integrity and ensuring the implementation of the design confining functions have been made. Based on the calculation results, it can be concluded that strength of the containment cylindrical part during a beyond design-basis accident cannot be ensured under parameters t (temperature) = 120°С, p (pressure) = 0.6 MPa.

Effect of pipe expansion on the redistribution of residual stresses after molding

2020 ◽  
Vol 10 (2) ◽  
pp. 122-126
Author(s):  
Nikolay A. Makhutov ◽  
◽  
Dmitry A. Neganov ◽  
Eugeny P. Studenov ◽  
◽  
...  
Keyword(s):  
Residual Stresses ◽  
Strain Distribution ◽  
Strain State ◽  
Stress And Strain ◽  
Cylindrical Shape ◽  
Pipe Material ◽  
Pipe Expansion ◽  
Calculation Results ◽  
Residual Deformations ◽  
Stress And Strain Distribution

In the factory, pipes for trunk oil and oil product pipelines are obtained by molding and welding. To ensure a cylindrical shape and reduce technological residual stresses, expansion technology is used. Pipe expansion causes a significant change in the values of residual deformations and stresses. The article presents both the calculation results and graphs regarding stress and strain distribution during bending of the stock and their redistribution after expansion. Based on the calculation results, the final total values of residual stresses and residual deformations caused by bending and expansion were stated to be important components of the stress-strain state observed in pipelines being operated under cyclic loading, as well as those used in assessing how degradation affects the ductility of the pipe material. These factors were concluded as being reasonably taken into account when performing verification calculations regarding long-running pipelines if, based on their diagnostics and analysis, their state does not meet modern strength requirements.

Design of Severe Accident Management Systems for Beyond Design Basis External Hazards at Paks NPP

2016 ◽  
Author(s):  
Tamás János Katona ◽  
András Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Management Systems ◽  
Basic Question ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Management ◽  
External Events

Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some mew systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented and design the new post-Fukushima measures that is outlined in the paper. The concept and practice is presented on several examples.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

Mass/Energy Release Influence to Thermal Hydraulics of Accident Localisation System at Ignalina NPP in Case of MDBA

2002 ◽  
Author(s):  
E. Urbonavicius ◽  
S. Rimkevicius
Keyword(s):  
Energy Release ◽  
Hydraulic Parameters ◽  
Thermal Hydraulics ◽  
Mass Energy ◽  
Design Basis ◽  
Calculated Mass ◽  
Design Basis Accident ◽  
Main Circulation ◽  
Maximum Design

The maximal loads on ALS structures could be reached during Maximum Design Basis Accident (MDBA). MDBA for Ignalina NPP is defined as guillotine rupture of the pressure header of main circulation pumps (MCP). In this paper the influence of calculated mass and energy release in case of MCP pressure header rupture on behavior of thermal hydraulic parameters in ALS compartments is presented. The performed analysis demonstrated the importance of mass/energy release for the calculated highest pressure in ALS. Nevertheless, the maximal calculated pressures in ALS compartments are below the design pressures even in case of most conservative mass/energy release through the break.

Simulation of Hydrogen Distribution in Ignalina NPP ALS Compartments During BDBA

2006 ◽  
Author(s):  
Egidijus Babilas ◽  
Egidijus Urbonavicius ◽  
Sigitas Rimkevicius
Keyword(s):  
Control System ◽  
Volume Concentration ◽  
Combustible Mixture ◽  
Hydrogen Distribution ◽  
Radiation Hazards ◽  
H2 Control ◽  
Design Basis ◽  
Design Basis Accident ◽  
Hydrogen Accumulation ◽  
Maximum Design

Accident Localisation System (ALS) of Ignalina NPP is a “pressure suppression” type confinement, which protects the population, employees and environment from the radiation hazards. According to the Safety Analysis Report for Ignalina NPP ∼110 m3 of hydrogen is released to ALS compartments during the Maximum Design Basis Accident. However in case of beyond design basis accident, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. If the volume concentration of hydrogen in the compartment reaches 4%, there is a possibility for a combustible mixture to appear. To prevent the possible hydrogen accumulation in the ALS of the Ignalina NPP during an accident the H2 control system is installed. The results of the performed analysis derived the places of the possible H2 accumulation in the ALS compartments during the transient processes and assessed the mixture combustibility in these places for a beyond design basis accident scenario. Such analysis of H2 distribution in the ALS of Ignalina NPP in case of BDBA was not performed before.

Ignalina accident localisation system response to maximum design basis accident

2003 ◽  
Vol 42 (2) ◽  
pp. 105-111 ◽  
Author(s):  
G. Dundulis ◽  
A. Kaliatka ◽  
S. Rimkevicius
Keyword(s):  
System Response ◽  
Design Basis ◽  
Design Basis Accident ◽  
Maximum Design

Accessibility Analysis for Staff After Accident of PWR Unit

2017 ◽  
Author(s):  
Ye Zhang ◽  
Xiaoxia Wang ◽  
Wei You ◽  
Zhuoran Li ◽  
Liying Zhang ◽  
...  
Keyword(s):  
Radioactive Material ◽  
Severe Accident ◽  
Accessibility Analysis ◽  
Dose Rates ◽  
Design Basis ◽  
Design Basis Accident

Severe accident has become one of the main directions of research since the crisis at Fukushima plants in Japan, including release of radioactive material, accessibility analysis for staff and evaluation of consequence. This paper, mainly for design basic accident and severe accident, makes calculation of migration and release of radioactive material after accident by considering the different building, also combine the on-site operation requirement of worker after accident, analyzes dose rates of typical zone for staff and evaluate the exposure caused by radioactive material. The main results of the paper supply reference and basis for person accessibility research after design basis accident and severe accident.

CANDU 6 Severe Accident Prevention and Mitigation Features

2012 ◽  
Author(s):  
Steven Ford ◽  
Boris Lekakh ◽  
Ed Choy ◽  
Kamal Verma ◽  
Sorin Ghelbereu
Keyword(s):  
Accident Prevention ◽  
Severe Accident ◽  
Multiple Failures ◽  
Core Damage ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Conditions

The CANDU 6 design includes features, both engineered and inherent, that act as barriers to prevent and mitigate severe accidents at progressive stages of a beyond design basis event such as that which occurred at Fukushima in March 2011. CANDU 6 has ample design margins including multiple layers of defense. Large inventories of water slow down any accident progression to severe accident conditions, even when multiple failures are assumed; giving operations staff more time to manage the event. Ongoing improvements to operating plants, and enhancements made to future evolutions of the CANDU design (including the Enhanced CANDU 6) improve upon these inherent features, further strengthening the CANDU 6 design to withstand severe core damage accidents.

Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion

2014 ◽  
Author(s):  
Genn Saji
Keyword(s):  
Hydrogen Generation ◽  
Hydrogen Gas ◽  
Severe Accident ◽  
Normal Operation ◽  
Water Radiolysis ◽  
Chain Reactions ◽  
Decay Heat ◽  
Severe Accidents ◽  
Heat Curve ◽  
Fukushima Daiichi

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1. When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters. Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.

scholarly journals Experimental research of the air permeability of concrete in the compressed zone of the containment for power units no. 1, 2 of the South-Ukrainian NPP

2021 ◽  
Vol 20 ◽  
pp. 39-50
Author(s):  
Ia. A. Zhygalov ◽  
◽  
V. V. Inyushev ◽  
V. O. Posokh ◽  
S. A. Vyzhva ◽  
...  
Keyword(s):  
Experimental Studies ◽  
Air Permeability ◽  
Complex Stress State ◽  
Air Pressure ◽  
Design Basis ◽  
Design Basis Accident ◽  
The Impact ◽  
Experimental Researches ◽  
Maximum Design

The determination of the air permeability of concrete in the compressed zone of NPP’s containment under conditions of excessive air pressure in the subshell space of the reactor compartment under a maximum design basis accident is necessary to confirm the localizing functions of the containment when extending the lifetime of power units. Approaches to determining the localizing ability (air permeability) of concrete in the compressed zone of the containment of NPPs with WWER-1000 type reactors under the impact of excessive air pressure under conditions of maximum design basis accident are discussed in the article. The designed testing installation for experimental studies of the air permeability of concrete by the method of stationary radial filtration is described and the results of experimental researches of the air permeability on the installation are presented for samples simulating the composition of concrete used under construction of the containment for Power Units No. 1, 2 (series B-302 and B-338, respectively) SD “South-Ukrainian NPP”. A comparative analysis of the results of abovementioned experimental researches with the results of experimental and theoretical work on the study of air permeability of the similar samples-imitators of concrete by the method of stationary linear filtration, as well as the results of measurements for samples of concrete sampled directly from the compressed zone of containment for Power Unit No. 2 of SD “South-Ukrainian NPP” was fulfilled. The results are explained by processes of compaction of the concrete structure under a complex stress state that occurs under specific hardening conditions and during longtime compression by pre-stressed reinforcing ropes, which takes place under conditions of real containment operation. It was concluded that the simulators made according to the currently accepted technology cannot provide a correct determination of the localizing ability of the NPP’s containment, since the conditions and time of containment concrete hardening, as well as the longtime constant compressive action of reinforcing ropes, cannot be correctly modeled in full under the formation of samples-imitators. The issue of the containment concrete samplesimitators creating in laboratory conditions requires additional study, improvement of technology and the development of new approaches with maximum physical modeling of the conditions characteristic of the operation of the NPP containment.

Sign in / Sign up

Export Citation Format

Share Document