Investigation of VVER-440 Confinement Response During DBA After Leaktightness Improvement

2020 ◽  
Author(s):  
Hovhannes Hovhannisyan ◽  
Tsolak Malakyan ◽  
Armen Amirjanyan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Active Safety ◽  
Main Interest ◽  
Design Basis ◽  
Active Safety Systems ◽  
Design Basis Accident ◽  
Safety Systems ◽  
Armenian Nuclear Power Plant

Abstract Based on Armenian Nuclear Power Plant (VVER-440/270) example the response of confinement to leaktightness improvement in case of design basis accident was investigated. The main interest of the investigation is to determine the influence of confinement leaktightness improvement on the reduction of available NPSH of active safety systems pumps (ECCS and spray systems). Results of analyses showed that confinement leaktightness improvement will lead to strong underpressure condition and endanger ECCS pumps operation due to cavitation. Hence, in parallel with confinement leaktightness improvement confinement vacuumization prevention means must be considered.

scholarly journals Loss of offsite power accident analysis in a VVER-1000/V446 nuclear power plant

2019 ◽  
Vol 34 (3) ◽  
pp. 231-237
Author(s):  
Mohsen Esfandiari ◽  
Gholamreza Jahanfarnia ◽  
Kamran Sepanloo ◽  
Ehsan Zarifi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Active Safety ◽  
Accident Analysis ◽  
Hydraulic Analysis ◽  
Active Safety Systems ◽  
Safety Systems ◽  
Coolant System ◽  
Relap5 Code

The aim of this study is to present a thermo-hydraulic analysis of the loss of offsite power accident in VVER-1000/V446 nuclear power plant using the RELAP5 code. Loss of offsite power accident would lead to the unavailability of major active safety systems, and that the safety criteria ensuring a secure operation of the nuclear power plant would be violated, resulting in core heat-up with possible core degradation. Therefore, the analysis and investigation of the plant, during this accident, is very important. For this purpose, different behaviors of major components in the primary and the secondary sides of the reactor coolant system are studied. In this paper, the reactor is simulated by using RELAP5 system code. The results show a reasonable agreement with the developed model and also with the Bushehr nuclear power plant final safety analysis reports.

scholarly journals Basic strategies of public protection in a nuclear power plant beyond—Design basis accident

2016 ◽  
Vol 2 (1) ◽  
pp. 24-29
Author(s):  
V.A. Kutkov ◽  
V.V. Tkachenko ◽  
S.P. Saakian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Design Basis ◽  
Design Basis Accident

The Analysis of AP1000 Beyond Design Basis DEDVI Accident

2013 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Heat Removal ◽  
The Core ◽  
Design Basis ◽  
Injection Line ◽  
Design Basis Accident ◽  
Core Cooling

Double ended break of direct vessel injection line (DEDVI) is the most typical small-break lost of coolant accident (LOCA) in AP 1000 nuclear power plant. This study simulated the DEDVI (without actuation of automatic depressurization system 1–3 stage valves, accumulators and passive residual heat removal heat exchanger) beyond design basis accident (BDBA) to validate the safety capability of AP1000 under such conditions. The results show that the core will be uncovered for about 863 seconds and then recovered by water after gravity injection from IRWST into the pressure vessel. The peak cladding temperature (PCT) goes up to 838.08°C, much lower than the limiting value 1204°C. This study confirms that in the DEDVI beyond design basis accident, the passive core cooling system (PXS) can effectually cool the core and preserve it integrate, and ensure the safety of AP 1000 nuclear power plant.

scholarly journals Radiation risk assessment for nuclear power plant staff in different emergency situations

2021 ◽  
Vol 66 (3) ◽  
pp. 378-384
Author(s):  
V. I. Orlovskaya ◽  
A. G. Trifonov
Keyword(s):  
Risk Assessment ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Radiation Risk ◽  
Internal Exposure ◽  
Plant Design ◽  
Emergency Situations ◽  
Design Basis ◽  
Design Basis Accident

The paper presents the results of radiation risk assessment for the staff of a nuclear power plant design during design basis accident (spent nuclear fuel assembly falling on fuel in reactor core or storage pool during refueling operations) and a beyond design basis accident (large leakage of the primary coolant with failure of the active part of the emergency cooling system and complete blackout for 24 h). The assessment is based on state-of-the-art radiation risk models from the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) and the International Commission on Radiological Protection (ICRP). The calculation of risk indicators for occupational exposure of NPP staff in emergency situations was carried out on the basis of data obtained using a computational module created in the COMSOL 5.6 multiphysics software, doses from a radioactive cloud and internal exposure due to inhalation for such radionuclides as 134Cs, 137Cs, 131I, 133I, 90Sr. A feature of this approach is the detailed consideration of the NPP industrial site infrastructure, which allows obtaining a more accurate assessment of the radionuclide air distribution and fallout.

Structural analysis of the CAREM-25 nuclear power plant subjected to the design basis accident and seismic loads

2017 ◽  
Vol 108 ◽  
pp. 42-56 ◽  
Author(s):  
Daniel Ambrosini ◽  
Ramón H. Codina ◽  
Oscar Curadelli ◽  
Carlos A. Martínez
Keyword(s):  
Structural Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Loads ◽  
Design Basis ◽  
Design Basis Accident

scholarly journals Basic strategies of public protection in a nuclear power plant beyond-design basis accident

2015 ◽  
Vol 2015 (4) ◽  
pp. 5-14 ◽  
Author(s):  
Vladimir Anatiolievich Kutkov ◽  
Valery Vasilievich Tkachenko
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Design Basis ◽  
Design Basis Accident

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

Study for Application of Safety Soft Control and Information System in Highly-Integrated Control Rooms of CPR1000 NPP

2013 ◽  
Author(s):  
Xuegang Zhang ◽  
Wei Liu ◽  
Hai Chang ◽  
Jianbo Wen ◽  
Yiqian Wu ◽  
...  
Keyword(s):  
Information System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
System Design ◽  
Nuclear Power ◽  
Integrated Control ◽  
Continuous Control ◽  
Design Basis ◽  
Control Rooms ◽  
Soft Control

For most of the newly built nuclear power plants, the computerized main control rooms (MCR) are adopted. The soft control, the typical feature of computerized Human-Interface System (HIS) in the computerized main control room and mediated by software rather than by direct physical connections, is comprised of safety and non-safety control interface which provides the operators with manual control for component-level, and allows both continuous control of plant process and discrete control of components in nuclear power plant. The safety soft control and information system (SSCIS) is used to give the safety commands to and check the immediate response of the safety process. This paper describes the application of the system design basis, functionality, communication, operation faceplate and system modes for SSCIS which is firstly introduced in CPR1000 nuclear power plant. The design criteria and basic design features of SSCIS is developed to be as the design basis of the design implementation. The ISG-04 ‘Highly-Integrated Control Rooms-Communications issues (HICRc)’ provides acceptable methods for addressing SSCIS communications in digital I&C system design. The NUREG0700 ‘Human-System Interface Design Review Guidelines’ is applied as reference for human factor engineering requirement in the SSCIS design. And the SSCIS design has also fully considered the possible customer usual practice.

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