scholarly journals Effectiveness and adverse effects of reactor coolant system depressurization strategy with various severe accident management guidance entry conditions for OPR1000

2014 ◽  
Vol 52 (5) ◽  
pp. 695-708 ◽  
Author(s):  
Seungwon Seo ◽  
Yongjae Lee ◽  
Seongnyeon Lee ◽  
Hwan-Yeol Kim ◽  
Sung Joong Kim
Keyword(s):  
Adverse Effects ◽  
Severe Accident ◽  
Reactor Coolant ◽  
Accident Management ◽  
Management Guidance ◽  
Coolant System

Preliminary Analysis of Effect of the Intentional Depressurization on Fission Product Behavior During TMLB’ Severe Accident

2009 ◽  
Author(s):  
Gaofeng Huang ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Fission Product ◽  
Reactor Vessel ◽  
Inert Gas ◽  
Severe Accident ◽  
Base Case ◽  
Reactor Coolant ◽  
Accident Management ◽  
Direct Containment Heating ◽  
Accident Sequences ◽  
Coolant System

It has been identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.

Experimental study on transport behavior of cesium iodide in the reactor coolant system under LWR severe accident conditions

2020 ◽  
Vol 57 (12) ◽  
pp. 1287-1296
Author(s):  
Naoya Miyahara ◽  
Shuhei Miwa ◽  
Mélany Gouëllo ◽  
Junpei Imoto ◽  
Naoki Horiguchi ◽  
...  
Keyword(s):  
Experimental Study ◽  
Cesium Iodide ◽  
Severe Accident ◽  
Reactor Coolant ◽  
Transport Behavior ◽  
Accident Conditions ◽  
Coolant System

The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

2012 ◽  
Vol 614-615 ◽  
pp. 626-631
Author(s):  
Chang Hong Peng ◽  
Ying Hao Yang
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Severe Accident ◽  
Frequency Model ◽  
Early Release ◽  
Accident Management ◽  
Management Guidance ◽  
The Mean ◽  
Accident Conditions ◽  
Thermal Hydraulic Analysis

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

Severe Accident Steady-State Modeling of Advanced Nuclear Power Plant Based on ASTEC Code

2017 ◽  
Author(s):  
Wei Song ◽  
Jiaxu Zuo ◽  
Yan Chen ◽  
Chaojun Li ◽  
Peng Zheng
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Cooling Systems ◽  
Accident Management ◽  
Management Guidance ◽  
Level 2 ◽  
Reactor Block

Severe accident is an attractive topic today for the nuclear power plant (NPP) safety. In the nuclear safety regulatory work, it is planned to build a full scale severe accident model for the advanced nuclear power plant of China to study the new designs of severe accident prevention and mitigation systems and strategies, and to further deploy the application on the level 2 PSA and severe accident management guidance. This paper firstly introduces the modeling tool, ASTEC, and then presents the progress of modeling work, which is mainly on the steady state modeling and regulation including reactor block, primary and secondary cooling systems, regulation systems etc. Last but not least, the work plan for the future is given.

Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear power plants

2008 ◽  
Vol 238 (4) ◽  
pp. 1093-1099 ◽  
Author(s):  
Thinh Nguyen ◽  
Raj Jaitly ◽  
Keith Dinnie ◽  
Ron Henry ◽  
Don Sinclair ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management ◽  
Management Guidance

scholarly journals VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions

1990 ◽  
Author(s):  
T.J. Heames ◽  
D.A. Williams ◽  
N.A. Johns ◽  
N.M. Chown ◽  
N.E. Bixler ◽  
...  
Keyword(s):  
Mechanistic Model ◽  
Severe Accident ◽  
Reactor Coolant ◽  
Accident Conditions ◽  
Coolant System

Assessment for Effect of RCS Depressurization and Reactor Cavity Flooding Using PSA Approach

2010 ◽  
Author(s):  
Changhong Peng ◽  
Ning Zhang ◽  
Pingping Liu
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Systematic Approach ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Past ◽  
Accident Management ◽  
Management Guidance

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will reduce the amount of radionuclides released to the environment during the accident. The plant studied is a three-loop pressurized water reactor (PWR) with large dry containment. The RCS depressurization and reactor cavity flooding can be used as an accident management strategy. Then, the decrease of LERF (Large and Early Release Frequency) is quantified using PSA approach. It can be found that strategy of RCS depressurization and reactor cavity flooding can mitigate the result of severe accident effectively.

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