Discussion on Application of Accident Sequence Precursor Analysis in China

2017 ◽  
Author(s):  
Jiang Hu ◽  
Guoxu Zhang ◽  
Yushu Zhang ◽  
Binbin Zhang ◽  
Zhi Xiao
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Operating Experience ◽  
Evaluation Process ◽  
Management Platform ◽  
Core Damage ◽  
Current State ◽  
Core Cooling ◽  
Damage Risk ◽  
Degraded Condition

An accident sequence precursor (ASP) is an initiating event or degraded condition that, when coupled with one or more postulated events, could result in inadequate core cooling or even severe reactor core damage. Through the systematical investigation and evaluation of nuclear power plant (NPP) operating experience, the ASP analysis could provide a comprehensive, risk-informed view of NPP operating experience, a measure for trending core damage risk, a partial validation of the current state of practice in risk assessment, and also a feedback to regulatory activities. This paper firstly gives a brief review on the ASP evaluation process of U.S. and German, and then introduces the current progresses on the establishment of ASP regulatory framework and the development of ASP management platform, which are carrying out by the related nuclear regulation agency in China. Also, this paper gives some insightful discussions on issues such as the ASP application and problems maybe faced.

Test Analysis of CAP1400 5cm Break Accident Based on the ACME Test Facility

2016 ◽  
Author(s):  
Sheng Zhu
Keyword(s):  
System Design ◽  
Nuclear Power ◽  
Reactor Core ◽  
Safety System ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water ◽  
Power Simulation ◽  
Passive Safety System ◽  
Core Cooling

CAP1400 is a large pressurized water reactor based on the passive safety conception. An ACME (Advanced Core-cooling Mechanism Experiment) facility has been designed and constructed in order to validate that the CAP1400 system design is acceptable to mitigate the loss of coolant accident (LOCA). The ACME test facility is an isotonic pressure, 1/3-scale height and 1/54.32-scale power simulation of the prototype CAP1400 nuclear power plant. It contains the main-loop system, passive safety system, secondary steam system and auxiliary system etc. The all of ACME test matrix including 5 kinds 21 cases .In this paper, the test results and the Realp5 prediction of the cold leg 5cm break accident of CAP1400 are compared and analyzed to briefly evaluate the ACME capability. Furthermore, 3 different types of 5cm cold leg break test cases are presented, and the transient process, system responses and key parameters tendency are analyzed based on the test. The results indicate that the passive safety system design can successfully combine to provide a continuous removal of core decay heat and the reactor core remains to be covered with considerable margin for the 3 different 5cm cold leg break accidents.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

The Use of Risk-Informed Inspection Approaches of Ukraine

2013 ◽  
Author(s):  
Bogdan Hryshchenko ◽  
Mykhailo Polianskyi ◽  
Anatoliy Nosovskyy ◽  
Oleksandr Sevbo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Experience ◽  
State Inspection ◽  
Core Damage ◽  
Unit Design ◽  
Low Efficiency ◽  
International Financial Corporation

Currently in Ukraine inspection activity is based on deterministic conservative principles, operation experience and expert appraisal of the inspector. Possibility and benefits of PSA are used with low efficiency. Results of the study conducted by IFC (International Financial Corporation) indicate the absence of a risk-informed approach in the practice of inspections in Ukraine. Also, according to the World Bank appraisal of the investment climate in Ukraine it should be concluded that until regulatory authorities begin implementation of a risk-informed approach in planning inspections in Ukraine random unscheduled events will dominate and won’t provide the goal of State Inspection. Information which obtained from the PSA helps to direct human and financial resources to the problems research that are the most important for safety, and to eliminate or reduce the requirements, which will reduce expenses of solution for significant issues. Inspection in the planning, preparation, implementation and evaluation of the results of which, in addition to deterministic estimates, operating experience and expertise evaluating risk are used is called the risk-informed inspection. The use of risk-informed approaches allows to: focus inspections on design and operational aspects, which have dominant influence on the safety of nuclear power plants (NPP); improve the schedule of inspections (recording of risk assessments in determining the scope, frequency, and type of inspection); and use an additional source of information on the systems and the components of power units, personnel availability. Applying a risk-informed approach to inspection, inspectors can focus primarily on systems that make the largest contribution to core damage frequency, failure of which leads to significant increase of CDF. Risk-informed approach allows to select the most important elements to test systems that will increase the efficiency and quality of inspections. Based on the above it can be concluded that the experience of inspector, his knowledge of the power unit design, the process, the mechanisms of failure of equipment and of accident running, the use of information on the importance of components and systems for the safety of nuclear power plants, obtained from PSA — an effective way to achieve the best results in improving safety.

Modification to Shield Tank Overpressure Protection in a CANDU Reactor for Beyond Design Basis Event

2019 ◽  
Author(s):  
Michael Huang ◽  
Khurram Khan ◽  
Ali Etedali-Zadeh ◽  
Jefferson Tse ◽  
Bing Li
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Lower Probability ◽  
Damage State ◽  
Core Damage ◽  
Design Basis ◽  
System Pressure ◽  
Severe Event ◽  
Large Heat

Abstract The Shield Tank and End Shield Cooling System in the CANDU reactor contains a large volume of light water surrounding the Calandria and circulates water to remove heat that arises from the reactor core and Moderator. In a beyond design basis event that results in a severe event, progression in the absence of mitigating cooling actions could result in a large heat load being transferred to the water inside the shield tank from the calandria wall causing shield tank failure due to over pressurization. Following the 2011 events at Fukushima Daiichi Nuclear Power Plant, the adequacy of system pressure relief was assessed against severe events. Emergency mitigating equipment tie-ins for water make-up will likely limit the core damage state and prevent the need to protect the shield tank. However, Shield Tank Overpressure Protection (STOP) has been installed against severe event conditions pursuant to the CANDU defense-in-depth safety philosophy. A larger open vent line has been installed at some CANDU units on the top of the shield tank outside containment. This design routes the vent piping high enough to preclude any venting under any operational configuration and discharges back into the containment through an existing spare penetration. Vent piping is designed as Nuclear Class 2 in accordance with ASME BPVC Section III. Assessment of stresses in the modification piping was also completed for BDBEs including for a lower probability seismic event, steam venting and corresponding higher pressure and temperature conditions.

Severe Accident Condition Modeling in PWR Environment: Creep Rupture Modeling

2013 ◽  
Author(s):  
Frederick W. Brust ◽  
R. Iyengar ◽  
M. Benson ◽  
Howard Rathbun
Keyword(s):  
High Temperatures ◽  
Nuclear Power ◽  
Reactor Core ◽  
Creep Rupture ◽  
Severe Accident ◽  
Time To Failure ◽  
Analysis Model ◽  
Pressurized Water ◽  
Condition Modeling ◽  
Core Cooling

A problem of interest in the nuclear power industry involves the response of pressurized water reactor (PWR) pressure boundary components under long-term station blackout (SBO) conditions. SBO is a particularly challenging event to nuclear safety, since all alternating current power required for core cooling is lost. If unmitigated, such a scenario will eventually lead to the reactor core being uncovered. Thermal-hydraulic (T-H), computational fluid dynamics, and structural combined creep/plasticity analyses of this scenario have been conducted and are presented here. In this severe accident scenario, high temperatures can occur, and impart this thermal energy to the surrounding structures, including the reactor vessel, nozzles, reactor coolant system (RCS) hot leg piping and S/G tubes. At such high temperatures and pressures, creep rupture of RCS piping and/or steam generator (S/G) tubes becomes possible. The intent of this paper is to present a finite element based analysis model that can be used to evaluate the time to failure of the nozzle-weld-pipe configuration.

Sensitivity Analyses of Steam Generator Tube Thickness on Induced-SGTR in SBO Accident

2013 ◽  
Vol 341-342 ◽  
pp. 1338-1341
Author(s):  
Jin Rong Qiu ◽  
Chang Hong Peng ◽  
Yun Guo
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Fission Products ◽  
Sensitivity Analyses ◽  
Steam Generator Tube ◽  
Core Damage ◽  
Reference Plant ◽  
Station Blackout ◽  
Level I ◽  
Damage Risk

For nuclear power plant, station black-out (SBO) is the events that contribute significantly to the level-I core damage risk. For an SBO, it is assumed that both the off-site power and on-site diesel generators fail to supply alternating-current power for the plant systems. SBO induced steam generator tube rupture (SGTR) is a concern because the steam generator (SG) tubes are parts of the reactor coolant pressure boundary and failure of the SG tubes may lead to fission products bypassing the containment. The SG tube integrity may be challenged by high temperature and high pressure conditions and may have a potential to fail due to creep rupture. This study focuses on the probability of SBO induced SGTR accidents under the station blackout (SBO) with RCS integrity, seal LOCA and steam relief valves remaining stuck open for the reference plant. At last, the sensitivity of the tube thick is studied.

scholarly journals Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-11 ◽  
Author(s):  
Gilberto Espinosa-Paredes ◽  
Raúl Camargo-Camargo ◽  
Alejandro Nuñez-Carrera
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Injection ◽  
Reactor Core ◽  
Cooling Water ◽  
Fission Products ◽  
Severe Accident ◽  
Core Damage ◽  
Laguna Verde

The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

The iROCS Approach to Mitigating Beyond-Design-Basis External Events

2016 ◽  
Author(s):  
Jaewhan Kim ◽  
Soo-Yong Park ◽  
Kwang-Il Ahn
Keyword(s):  
Coping Strategies ◽  
Control Function ◽  
Reactor Core ◽  
Heat Removal ◽  
Monitoring And Control ◽  
Plant Damage ◽  
Core Damage ◽  
Design Basis ◽  
External Events ◽  
Core Cooling

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.

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Alexander Vasiliev ◽  
Juri Stuckert
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Post Test ◽  
Core Cooling ◽  
Modeling Code ◽  
Code Quality

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

Emergency Core Cooling System Sump Strainer Fibre Bypass Testing and Measurement

2016 ◽  
Author(s):  
Qingwu Cheng ◽  
Harry Adams ◽  
Metin Yetisir
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Regulatory Requirements ◽  
Pressurized Water ◽  
Load Pressure ◽  
Analysis Techniques ◽  
Core Cooling

The potential of losing post-Loss Of Coolant Accident (LOCA) recirculation capability due to debris blockage of Emergency Core Cooling (ECC) strainers resulted in early replacements of ECC strainers in most nuclear power plants. To validate the performance of ECC strainers, extensive testing representing plant conditions is required. Such testing programs include thin-bed and full debris load pressure drop tests, fibre bypass tests and chemical effects tests. Multiple testing loops and state-of-the-art analysis techniques have provided in-depth understanding of sump strainer performance and the effect of chemical precipitation on debris bed head loss. ECC strainers typically use perforated plates as filtering surfaces with 1.6 to 2.5 mm holes and 35 to 40% open area, allowing some particulates and fibres to pass through the strainer filtering surfaces. Recently, the bypassed fibrous debris has been identified as a potential safety concern due to its possible deposition in the reactor core and blocking of flow into fuel assemblies. In some cases, the amount of fibre that is specified as allowed to enter a reactor core is only 15 g per fuel assembly for pressurized water reactors. Characterization and quantification of bypassed fibre debris for nuclear power plants are needed to address regulatory requirements. Testing methodology and analysis techniques to address regulatory requirements and concerns are presented in this paper. In particular, a newly developed technique is presented to address debris bypass quantification.

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