Passive Safety Systems of Advanced Nuclear Power Plant: AP1000

2010 ◽  
Author(s):  
Guohua Yan ◽  
Chen Ye
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Cooling System ◽  
Heat Removal ◽  
Injection System ◽  
Passive Safety ◽  
Human Errors ◽  
Power Sources ◽  
Passive Safety Systems ◽  
Safety Systems

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout

2014 ◽  
Author(s):  
Jie Zou ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Safety System ◽  
Injection System ◽  
Passive Safety ◽  
System Failure ◽  
Decay Heat ◽  
Passive Safety Systems ◽  
Station Blackout ◽  
Safety Systems

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.

Development of Passive Cooling System Models for the Assessment of the Long Term Cooling Capability in the Liquid Metal Reactor

2004 ◽  
Author(s):  
W. P. Chang ◽  
K. S. Ha ◽  
H. Y. Jeong ◽  
S. Heo ◽  
Y. B. Lee
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Computer Code ◽  
Passive Safety ◽  
Primary Concern ◽  
Metallic Fuel ◽  
Passive Safety Systems ◽  
Operator Action ◽  
Safety Systems

This study has been carried out to assess the decay heat removal capability of the passive safety systems adopted in a conceptual design of the 600 MW(e), sodium cooled, metallic fuel loaded KALIMER. The applicability of the PVCS, which used to be the only passive safety system for KALIMER-150, is limited to a reactor capacity of 1,000 MW(t) or less. Another passive loop, PDRC, is conceptualized in order to overcome the limit as the KALIMER capacity scales up from the current 150 MW(e) to 600 MW(e). The safety analysis computer code, SSC-K, currently used for KALIMER is not capable of simulating such passive systems. With this concern, the PVCS and PDRC models are developed and they are coupled with the SSC-K for a long-term cooling assessment. The present paper thus presents the analysis results of the ULOHS using these models along with their brief introductions. The primary concern of the analyses is focused on the inherent safety as well as the system’s integrity for 72 hours without any operator action during the event.

iB1350: A Generation III.7 Reactor After the Fukushima Daiichi Accident

2016 ◽  
Author(s):  
Takashi Sato ◽  
Keiji Matsumoto ◽  
Kenji Hosomi ◽  
Keisuke Taguchi
Keyword(s):  
Cooling System ◽  
Boiling Water ◽  
Passive Safety ◽  
Active Safety ◽  
Water Reactor ◽  
Airplane Crash ◽  
Passive Safety Systems ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

AP1000® Passive Cooling Containment Analysis of a Double-Ended LBLOCA With a 3D Gothic Model

2018 ◽  
Author(s):  
Samanta Estevez-Albuja ◽  
Gonzalo Jimenez ◽  
Kevin Fernández-Cosials ◽  
César Queral ◽  
Zuriñe Goñi
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Water Tank ◽  
Local Pressure ◽  
Water Storage Tank ◽  
Passive Safety Systems ◽  
Containment Model ◽  
Loss Of Coolant Accident ◽  
Ac Power ◽  
Safety Systems

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat sink without the need of AC power to work properly during normal or accidental conditions. To evaluate its performance, the AP1000 PCS has been usually modeled with a Lumped Parameters (LP) approach, coupled with another LP model of the steel containment vessel to simulate the accidental sequences within the containment building. However, a 3D simulation, feasible and motivated by the current computational capabilities, may be able to produce more detailed and accurate results. In this paper, the development and verification of an integral AP1000® 3D GOTHIC containment model, taking into account the shield building, is briefly presented. The model includes all compartments inside the metallic containment liner and the external shield building. Passive safety systems, such as the In-containment Refueling Water Storage Tank (IRWST) with the Passive Residual Heat Removal (PRHR) heat exchanger and the Automatic Depressurization System (ADS), as well as the PCS, are included in the model. The model is tested against a cold leg Double Ended Guillotine Break Large Break Loss of Coolant Accident (DEGB LBLOCA) sequence, taking as a conservative assumption that the PCS water tank is not available during the sequence. The results show a pressure and temperature increase in the containment in consonance with the current literature, but providing a greater detail of the local pressure and temperature in all compartments.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome BWR Reactor KERENA: Part II

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 109 ◽  
Author(s):  
René Manthey ◽  
Frances Viereckl ◽  
Amirhosein Moonesi Shabestary ◽  
Yu Zhang ◽  
Wei Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.

Dependency Analysis of Critical Parameters Influencing the Reliability of Thermal-Hydraulic Nuclear Passive Systems

2017 ◽  
Author(s):  
Samuel Abiodun Olatubosun ◽  
Zhijian Zhang
Keyword(s):  
Nuclear Power ◽  
Critical Parameters ◽  
Passive Safety ◽  
Distribution Analysis ◽  
Passive Systems ◽  
Research Issues ◽  
Passive Safety Systems ◽  
Influencing Parameters ◽  
Safety Systems ◽  
Nuclear Applications

The deployment of passive safety systems in nuclear applications especially in advanced nuclear power reactors (both evolutionary and innovative designs) is on the rise. This can be linked to the simplicity, economic and less dependence on human interventions attributes of those passive systems. The reliability of nuclear passive systems especially the thermal-hydraulic ones is influenced by parameters which are interdependent in reality. As a result, the need to critically consider the synergetic effects of determinants of reliability of the thermal-hydraulic nuclear passive systems is of utmost importance. Reliability methodologies are now being modified by factoring the dependency nature of those determinants into reliability analysis to obtain more realistic and accurate results. This paper thus focused on the introduction of more influencing factors in parameters dependency consideration of phenomenological reliability using multivariate distribution analysis. A passively water cooled steam generator was used to demonstrate the interdependency effects of some selected critical parameters. The results obtained justified the need for considering the dependency effects of these parameters influencing the reliability of thermal-hydraulic passive systems. In addition, the research issues on dependency consideration of influencing parameters in evaluation of reliability of these nuclear passive systems were also discussed.

scholarly journals Onshore Nuclear Power Plant Concept with Enhanced Passive Safety System

2019 ◽  
Vol 4 (6) ◽  
pp. 155-159
Author(s):  
A.H.M. Iftekharul Ferdous ◽  
T. H. M Sumon Rashid ◽  
Md Asaduzzaman Shobug ◽  
Tanveer Ahmed ◽  
Nitol Kumar Dutta
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Source ◽  
Passive Safety ◽  
Electric Power Supply ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Safety Systems

Bangladesh is a developing country and it’s increasing economy can be maintained by providing sufficient amount of electric power supply. Therefore government is initiating Rooppur nuclear power project is one of them which is needed to be sited beside a vast amount of water source, lowest populated area and away from the locality to reduce the damage caused by any nuclear accidents. In this thesis paper we have shown that, the the dangers of residing errors of Rooppur nuclear power plant and give a proposal to go for onshore nuclear power plant in Bangladesh with two proposed designs of passive safety systems PSS-I & PSS-II. These systems will give safety to the power plants in the case of plant blackout during accidents.

scholarly journals Experiment for Justification the Reliability of Passive Safety System in NPP

2018 ◽  
Vol 3 (3) ◽  
pp. 1
Author(s):  
D.S. Samokhin ◽  
Mohammad Alslman ◽  
A. D. Vostrilova ◽  
O.Yu. Kochnov
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Critical Issue ◽  
Nuclear Industry ◽  
Passive Safety ◽  
Mathematical Methods ◽  
Mean Time Between Failures ◽  
Passive Safety Systems ◽  
Passive Safety System ◽  
Safety Systems

This article gives an overview of the formation of the global nuclear industry, highlighted a critical issue of ensuring safe operation of nuclear power systems in modern projects. Considering the use of passive safety systems in the design of a nuclear power plant, and discussed the different mathematical methods for assessing the reliability of passive systems. Also it considers the possibility of finding the mean time between failures, using these methods to assess the reliability of passive safety systems.

scholarly journals Modelling of Passive Heat Removal Systems: A Review with Reference to the Framatome KERENA BWR Reactor: Part I

Energies ◽  
2019 ◽  
Vol 13 (1) ◽  
pp. 35 ◽  
Author(s):  
Amirhosein Moonesi Shabestary ◽  
Frances Viereckl ◽  
Yu Zhang ◽  
Rene Manthey ◽  
Dirk Lucas ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gravitational Force ◽  
Heat Removal ◽  
Dimensional System ◽  
Thermal Gradients ◽  
Two Phase ◽  
Passive Safety Systems ◽  
External Power ◽  
Safety Systems

Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modelling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I deals with numerical and experimental methods for modelling of condensation inside the emergency condensers and on the containment cooling condenser while part II deals with boiling and two-phase flow instabilities.

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