A COMPARATIVE ASSESSMENT OF PASSIVE SAFETY SYSTEMS IN INTEGRAL PRESSURIZED WATER REACTOR TYPE SMALL MODULAR REACTOR

2019 ◽  
Vol 2019.27 (0) ◽  
pp. 1872
Author(s):  
Yi Mi ◽  
Chireuding Zeliang ◽  
Akira Tokuhiro ◽  
Lixuan Lu
Keyword(s):  
Comparative Assessment ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Passive Safety Systems ◽  
Small Modular Reactor ◽  
Reactor Type ◽  
Safety Systems

Dynamic PRA Based on System Codes Coupling for Passive Safety System in Integral Pressurized Water Reactor

2020 ◽  
Author(s):  
Yi Mi ◽  
Akira Tokuhiro
Keyword(s):  
Natural Circulation ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Active Systems ◽  
Water Reactor ◽  
System Parameters ◽  
Event Ordering ◽  
Passive Safety Systems ◽  
The Difference

Abstract An integral Pressurized Water Reactor (iPWR) type SMR was studied featuring Passive Safety Systems (PSSs). Different from active systems, PSSs are easily influenced by system parameters referred to as phenomenological factors such as heat loss, flow friction, oxidation, non-condensable gases and void fraction due to the low driving force of natural circulation. The system parameters also contribute to the uncertainty and dependency of PSS leading to the system unreliability. Thus, efforts are made to improve the reliability of PSS. A classical Probabilistic Risk Assessment (PRA) model describing active systems does not consider time evolution nor event ordering for PSS that dynamic PRA can accommodate. Here we developed and realized coupling between LabVIEW and CAFTA. Isolation Condenser System (ICS) is taken as the benchmark system due to the simple design in single phase without phase change phenomena in order to mainly remove decay heat and secondarily depressurize the reactor pressure vessel (RPV). A classical PRA model of ICS using CAFTA is coupled with real-time simulation of primary loop and ICS in LabVIEW, leading to a dynamic simulation result. The difference in failure probability using dynamic versus classical PRA revealed that for one there are more component demands with different event ordering, such that improved PSS reliability in the iPWR-type SMR designs is possible.

LOCA Break Spectrum Analysis of the Westinghouse Small Modular Reactor

2013 ◽  
Author(s):  
Vefa N. Kucukboyaci ◽  
Jun Liao
Keyword(s):  
Spectrum Analysis ◽  
Evaluation Model ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Passive Safety ◽  
Plant Design ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Small Modular Reactor ◽  
Safety Systems

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17×17 fuel assembly design used in the AP1000® reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A break spectrum analysis on the Westinghouse SMR LOCA has been performed to investigate the performance of the SMR passive cooling. The break type includes both the double-ended guillotine (DEG) break and the split break with the break size ranging from 0.5 inch to the diameter of direct vessel injection (DVI) line. The break spectrum analysis was performed using the WCOBRA/TRAC-TF2 code, which is designed to simulate PWR LOCA events from the smallest break size to the largest break size. The break spectrum analysis demonstrates that excellent performance of the passive safety system of the Westinghouse SMR in variable LOCA conditions. The study is also a necessary step to develop an evaluation model for the analysis of design basis LOCA accident.

Film Boiling Behavior of a Pressurized Water Reactor Type Fuel Bundle

1981 ◽  
Vol 52 (1) ◽  
pp. 86-99 ◽  
Author(s):  
F. S. Gunnerson ◽  
D. T. Sparks ◽  
D. K. Kerwin
Keyword(s):  
Film Boiling ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Reactor Type ◽  
Type Fuel

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.

scholarly journals Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

2017 ◽  
Vol 2017 ◽  
pp. 1-16 ◽  
Author(s):  
Siniša Šadek ◽  
Davor Grgić ◽  
Zdenko Šimić
Keyword(s):  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
European Code ◽  
Station Blackout ◽  
Computer Codes ◽  
Safety Systems ◽  
Thermal Hydraulic Analysis

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

scholarly journals Role of Passive Safety Systems in Chinese Nuclear Power Development

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
X. Cheng ◽  
Y. H. Yang ◽  
Y. Ouyang ◽  
H. X. Miao
Keyword(s):  
Nuclear Power ◽  
Passive Safety ◽  
Pressurized Water Reactors ◽  
Chinese Government ◽  
Power Development ◽  
Pressurized Water ◽  
Passive Safety Systems ◽  
Nuclear Power Development ◽  
Water Reactors ◽  
Safety Systems

Passive safety systems have been widely applied to advanced water-cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near-term and medium-term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large-scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

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