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

Author(s):  
Liu Yu ◽  
Li Wenjing ◽  
Yu Yun

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.

2021 ◽  
Author(s):  
Yu Liu ◽  
Cong Wang ◽  
Jing Liu ◽  
Heng Gao

Abstract Fukushima accident is a tragic case that extreme external events (tsunami and seismic) threaten the safety of NPP and lead to large uncontrollable release of radioactive materials. Therefore, how to perform the study on external events Level 2 PSA has become a crucial problem needed to be considered deeply. In this study, the main purpose is to describe an external events Level 2 PSA approach and apply it for a III generation PWR in China. This study consists of the following factors: (1) Screening out external events. (2) Focusing on the key technical elements of Level 2 PSA with special considerations on the characteristics of different external events. (3) Building a detailed Level 2 PSA model for the screened events with software RiskSpectrum and obtaining the quantitative results of large release frequency (LRF). (4) According to the analysis results, finding out the risk weaknesses and summarizing recommended proposals.


Author(s):  
Robert J. Lutz ◽  
Robert P. Prior

The accident at the three reactor units at Fukushima Daiichi showed weaknesses in the plant coping capability for beyond design basis accidents caused by extreme external events. The weaknesses included plant design features, accident management procedures and guidance, and offsite emergency response. As a result, significant changes to plant coping capability have been made to light water reactors worldwide to enhance the coping capabilities for beyond design basis accidents. However, the response in the United States has been significantly different from that in Europe in a number of ways. In the United States, the regulator and the industry convened separate expert panels to review the Fukushima accident and make recommendations for enhancements. On the regulatory side, a series of three Orders were issued and that required the implementation of certain enhancements (Mitigation strategies, hardened vents for certain BWRs, spent fuel pool level indication) to ensure adequate protection for the health and safety of the public. Other enhancements were subject to the “Backfit Rule” which requires that changes to regulatory requirements be shown to be cost beneficial using accepted methodologies. Simultaneously, the industry took independent steps to develop a diverse and flexible coping strategies (known as FLEX) and other enhancements. The focus in the United States was clearly on enhancements to guarantee continued core, containment and spent fuel pool cooling in the event of beyond design basis accidents, particularly those resulting from extreme external events. In Europe, the regulatory agencies ordered the development and completion of “Stress Tests” for each reactor site. These Stress Tests were focused on identifying the capability of the plant and its staff to respond to increasingly severe external events. The Stress Tests not only examined the ability to maintain core, containment and spent fuel pool cooling but also the ability to mitigate the consequences of accidents that progress to core damage (i.e., a severe accident). Regulatory requirements were then issued by the national regulators that addressed the weaknesses identified from the Stress Tests. While many of the enhancements to the plant coping capability were similar to those in the United States, significant hardware enhancements were also required to reduce the consequences of core damage accidents including hydrogen control and containment filtered venting. Finally, most European regulators also include severe accident management guidance (SAMG) as a regulatory requirement. In the United States, SAMG will be maintained as a voluntary industry commitment that is subject to regulatory oversight review.


Author(s):  
Tamás János Katona ◽  
András Vilimi

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.


Author(s):  
Steven Ford ◽  
Boris Lekakh ◽  
Ed Choy ◽  
Kamal Verma ◽  
Sorin Ghelbereu

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.


Author(s):  
Robert J. Lutz ◽  
James Lynde ◽  
Steven Pierson

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.


2015 ◽  
pp. 3-8
Author(s):  
I. Bilodid ◽  
J. Duspiva

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.


Author(s):  
Jaewhan Kim ◽  
Soo-Yong Park ◽  
Kwang-Il Ahn

An extended loss of all electric power occurred at the Fukushima Dai-ichi nuclear power plant by a large earthquake and subsequent tsunami. This event led to a loss of reactor core cooling and containment integrity functions at several units of the site, ultimately resulting in large release of radioactive materials into the environment. In order to cope with beyond-design-basis external events (BDBEEs), this study proposes the iROCS (integrated, RObust Coping Strategies) approach. The iROCS approach is characterized by classification of various plant damage conditions (PDCs) that might be impacted by BDBEEs and corresponding integrated coping strategies for each of PDCs, aiming to maintain and restore core cooling (i.e., to prevent core damage) and to maintain the integrity of the reactor pressure vessel if it is judged that core damage may not be preventable in view of plant conditions. The plant damage conditions considered in the iROCS approach include combinations of the following conditions of the critical safety functions: (1) an extended loss of AC power, (2) an extended loss of DC power (loss of the monitoring and control function at control rooms), (3) a loss of RCS inventory, and (4) a loss of secondary heat removal. From a case study for an extreme damage condition, it is shown that candidate accident management strategies should be evaluated from the viewpoint of effectiveness and feasibility against extreme damage conditions of the site and accident scenarios of the plant.


2021 ◽  
Vol 2 (4) ◽  
pp. 398-411
Author(s):  
Jinho Song

Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of fishery products, discharge of radioactive water to the ocean, status of decommissioning, and needs for long-term monitoring of the environment are discussed.


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