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.

Study on Methodology and Application of Seismic-Induced Flood Level 2 PSA for PWR in China

2020 ◽  
Author(s):  
Liu Yu ◽  
Li Wenjing ◽  
Yu Yun
Keyword(s):  
Severe Accident ◽  
Key Factors ◽  
Fukushima Accident ◽  
Core Damage ◽  
Design Basis ◽  
Flood Level ◽  
Crucial Problem ◽  
External Events ◽  
Flood Characteristics ◽  
Level 2

Abstract Level 2 (L2) PSA is focused on the severe accident phenomenon, progression and source terms release to generally evaluate the containment response after core damage takes place. Fukushima accident was caused by seismic and tsunami which are beyond design basis. It indicates that evaluating the risk for extremely external hazards is vitally important. Therefore, how to perform the study on L2 PSA for external events (especially seismic and flood) has become a crucial problem needed to be considered deeply for both regulators and operators. In this paper, the methodology of flood and seismic-induced flood L2 PSA was developed and applied for a Gen III NPP in China. The key factors include: (1) Focusing on crucial elements of L2 PSA in view of seismic and flood characteristics, including PSA interfaces, design features, severe accident phenomenon and progression, containment performance analysis, etc. (2) Building integrated internal flood and seismic L2 PSA models. (3) Developing an analytical method to evaluate seismic-induced flood L2 PSA.

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 Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Use of PRA in the Design of the Westinghouse AP1000 Plant

2009 ◽  
Author(s):  
Robert J. Lutz ◽  
James H. Scobel ◽  
Richard G. Anderson ◽  
Terry Schulz
Keyword(s):  
Cooling Water ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Pressurized Water ◽  
Core Damage ◽  
Service Water ◽  
Containment Shell ◽  
Water Reactors ◽  
Accident Scenarios

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.

Methodology for Calculating Minor Radioactive Releases From VVER 1000 Using TRACE Code

2021 ◽  
Vol 7 (2) ◽  
Author(s):  
M. Ruscak ◽  
G. Mazzini ◽  
A. Dambrosio ◽  
A. Musa
Keyword(s):  
Power Plant ◽  
Severe Accident ◽  
Safety Requirements ◽  
Design Basis ◽  
Long Duration ◽  
Wide Range ◽  
Reduce Risk ◽  
Fukushima Daiichi ◽  
Accident Conditions ◽  
Action Levels

Abstract After the Fukushima DAIICHI accident, new safety requirements were imposed in order to reduce risk of severe accident. One of the principles that have been adopted is the introduction of emergency action levels resulting from the expected consequences. They cover a wide range of component and system malfunctions resulting in emergency, incident, and/or accident conditions. To evaluate those emergency action levels, thermal hydraulic (TH) analyses simulating these malfunction/incident/accident conditions are required. This paper describes the simulation of a real operational incident scenario using a standard thermal hydraulic model of the power plant in the TRACE code that was originally intended for simulation of design basis accidents such as large break coolant accident or loss of flow accident. Special attention was given to the methodology, addressing a long duration of an incident with corrective actions of the operators, and to computational issues leading to model modifications caused by a long duration of the incident along with the necessary conservatisms in the estimated results of the simulated radioactivity release.

Fundamental Experiments on Local Hydrogen Behaviors Under Severe Accident Conditions in a Rectangular Subcompartment of Nuclear Power Plant

2002 ◽  
Author(s):  
Jung-Jae Lee ◽  
Un-Jang Lee ◽  
Goon-Cherl Park
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Hydrogen Concentration ◽  
Concentration Distribution ◽  
Nuclear Power ◽  
Experimental Results ◽  
Severe Accident ◽  
Severe Accidents ◽  
Helium Gas ◽  
Accident Conditions

During severe accidents in nuclear power plant (NPP), the limit of local hydrogen concentration is regulated to ensure the integrity of containment. In this study hydrogen mixing experiments were conducted to investigate the hydrogen concentration distribution in a subcompartment of NPP. The mixing compartment is a vertical rectangular type with the dimension of 1×1×1.5 m3. The helium gas was used as a simulant of hydrogen. The goals of this study are to understand local hydrogen mixing phenomena and to examine the effects of the wall condensing, of the existence of obstacle and of the containment spray operation on local hydrogen concentration in NPP. Experimental results showed that the hydrogen might be locally accumulated in the subcompartment and the local hydrogen concentration could instantaneously rise during the spray operation.

FLEX Loss of Instrumentation Guidance for PWRs Enhances Severe Accident Diagnostics

2016 ◽  
Author(s):  
Robert J. Lutz ◽  
James Lynde ◽  
Steven Pierson
Keyword(s):  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Spent Fuel ◽  
Pressurized Water ◽  
The Core ◽  
Design Basis ◽  
External Events ◽  
Accident Conditions ◽  
Specific Implementation ◽  
Support Guidelines

The industry response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 is based on a set of Diverse and Flexible Coping Strategies (commonly referred to as FLEX) for beyond design basis external events as described in NEI 12-06. The Pressurized Water Reactors Owners Group (PWROG) developed generic guidance for response to these Beyond Design Basis External Events (BDBEE), called FLEX Support Guidelines (FSGs). These guidelines are referenced from the plant Emergency Operating Procedures (EOPs) when it is determined that an event exhibits certain beyond design basis characteristics such as an Extended Loss of all AC Power (ELAP). These generic FLEX Support Guidelines provide a uniform basis for all PWRs to implement the FLEX guidance in NEI 12-06 that was endorsed by the NRC to maintain core, containment and spent fuel cooling. The PWROG generic FSGs include guidance in FSG-7, “Loss of Vital Instrumentation or Control Power” for obtaining information for key plant parameters in an ELAP event. The key parameters were selected based on industry guidance and plant specific implementation. This set of key parameters will allow the licensed operators to have vital instrumentation to safely shutdown the core and maintain the core in a shutdown condition, including core, containment and spent fuel pool cooling. These parameters are used in the EOPs as well as the FSGs that are designed to mitigate a beyond design basis event. The requirements of NEI 12-06, as implemented through the FSGs, enhance both availability and reliability of instrumentation by requiring diverse methods of providing DC power for instrumentation and control as well as protection of instrumentation from the beyond design basis event. The subsequent implementation of this guidance at the Byron Station has proven to also be beneficial for diagnosis of severe accident conditions (where core cooling could not be maintained). The same parameter values that are needed to verify core, containment and spent fuel cooling prior to core damage are also needed to diagnose severe accident conditions. Guidance provided within FSG-7, as implemented at the Byron Station, contains several layers of diverse methods to obtain parametric values for key variables that can be especially useful when the environmental qualification is exceeded for the primary instrumentation that provides this information. The methods range from the use of self-powered portable monitoring equipment to the use of local mechanical instrumentation. The FSG-7 guidance is referenced from the Byron Severe Accident Management Guidance (SAMG) to either obtain parameter information during a severe accident or to validate the information that is available from the primary instrumentation.

Innovative Practice of Severe Accidents Measures in CAP1400

2017 ◽  
Author(s):  
Likai Fang ◽  
Xin Liu ◽  
Guobao Shi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Power Sources ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

scholarly journals Analysis of Fuel Criticality during Severe Accidents

2015 ◽  
pp. 3-8
Author(s):  
I. Bilodid ◽  
J. Duspiva
Keyword(s):  
Severe Accident ◽  
Stress Tests ◽  
Fukushima Accident ◽  
Management Guidelines ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Management ◽  
The World ◽  
Criticality Safety

Interest in the analysis of beyond design basis accidents, involving a combination of several failures with fuel damage, has increased throughout the world after the Fukushima accident. Stress tests were performed at NPPs, and development of severe accident management guidelines was started. These activities necessitated calculations to analyze the probability of beyond design basis accidents and assess their initiating events and consequences. One of the aspects in analysis of beyond design basis accidents is to determine the potential for re-criticality during such accidents. The paper provides results of some criticality safety calculations for VVER reactors performed, in particular, by ÚJV Řež and SSTC NRS experts. It is shown how criticality can occur in different severe accident phases.

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