scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Study of Station Blackout Accident Process in Taishan Nuclear Power Plant

2013 ◽  
Author(s):  
Yu Li ◽  
Huiyong Zhang ◽  
Yehong Liao ◽  
Jiming Lin ◽  
Dekui Zhan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Final State ◽  
Primary Coolant ◽  
Station Blackout ◽  
Full Power ◽  
Diesel Generators

According to the design features of the nuclear power plant in Taishan, a station blackout (SBO) at full power is a complex sequence, induced by the loss of offsite power (LOOP) combined with the loss of the emergency diesel generators (EDGs). This paper shall be performed the deterministic safety analysis of SBO without PSA study, and it mainly analyzes an overheating event of primary coolant resulting in the temporary unavailability of the steam generators (SGs) feeding systems during the station blackout. The analysis of this accident is carried out using the CATHARE2 thermal hydraulic code. The analysis pointed out that the final state can be reached after the operator performs the main manual actions such as startup of the SBO diesel generators from the main control room; startup of the two ASG pumps of trains 1 and 4; Local opening of the ASG header downstream the ASG pumps to enable the two power supplied ASG pumps to feed all the four SGs; initiation of a manual cooldown via the VDA in order to ensure the long term protection of the RCP pumps seals with respect to the thermal and mechanical loads. It corresponds to the achievement of a stable heat removal conditions by the emergency feedwater system (ASG) and the atmospheric steam dump system (VDA).

Study on the SBO Coping Capability for Qinshan Phase II Unit 3

2013 ◽  
Author(s):  
Jian Deng ◽  
Bin Chen ◽  
Chunrui Deng ◽  
Ming Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Event Tree ◽  
Natural Event ◽  
Core Damage ◽  
Battery Capacity ◽  
Station Blackout

The accident in Japan on March 2011 was caused by a natural event (i.e., earthquake, and the induced tsunami) which was far more severe than the design basis for the Fukushima Dai-ichi nuclear power plant. The reactor cores of Dai-ichi unit 1 through 3 have been degraded due to the long term station blackout (SBO). Actually, a nuclear power plant is able to prevent the core damage despite the SBO event occurs, called “coping capability” or “coping time”. In this paper, the SBO coping capability is assessed for Qinshan II unit 3 based on the following: 1) reactor coolant inventory, 2) condensate inventory for decay heat removal, 3) class 1E battery capacity, 4) compressed air, 5) effects of loss of ventilation, 6) containment isolation. As a result, the SBO event tree is suggested to be updated and a few vulnerabilities have been identified and some modifications are proposed.

scholarly journals Seismic safety evaluation during site selection for the nuclear power plants in Bangladesh

2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Heat Removal ◽  
Good Time ◽  
Diagnostics System ◽  
Seismic Impacts

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

2021 ◽  
Author(s):  
Janos Bodi ◽  
Alexander Ponomarev ◽  
Evaldas Bubelis ◽  
Konstantin Mikityuk
Keyword(s):  
Power Plant ◽  
Fast Reactor ◽  
Heat Removal ◽  
Current Paper ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Station Blackout ◽  
Safety Assessments ◽  
The Individual

Abstract As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Scaling Evaluation of the Two-Phase Natural Circulation

2012 ◽  
Author(s):  
Han Wang ◽  
Yuquan Li
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Characteristic Time ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Scaling Analysis ◽  
Top Down ◽  
Two Phase ◽  
Height Scale ◽  
Full Pressure

This paper presented the scaling evaluation of the two-phase natural circulation process between an assumed nuclear power plant and three test facilities with full pressure simulation and three different height scales, which were 1:2, 1:3 and 1:4. The Hierarchical Two-Tiered Scaling (H2TS) Methodology was adopted. By top-down scaling analysis, several characteristic time ratios were obtained, and then the calculation method of the scaling distortion were investigated. It has been found that the dominant processes in two-phase natural circulation can be well preserved no matter what the height scale is.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

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