Development of a thermal-hydraulic analysis code for primary system of an integrated small modular pressurized water reactor

2021 ◽  
Vol 152 ◽  
pp. 108019
Author(s):  
Dongqing Wang ◽  
Huandong Chen ◽  
Jiayue Chen ◽  
Dekui Zhan ◽  
Shaoxiong Xia
Keyword(s):  
Primary System ◽  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Hydraulic Analysis

Thermal hydraulic analysis for grid supported pressurized water reactor cores

2009 ◽  
Vol 239 (8) ◽  
pp. 1442-1460 ◽  
Author(s):  
C. Shuffler ◽  
J. Trant ◽  
J. Malen ◽  
N. Todreas
Keyword(s):  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Cores ◽  
Thermal Hydraulic Analysis

scholarly journals A RELAP5 model for the thermal-hydraulic analysis of a typical pressurized water reactor

2010 ◽  
Vol 14 (1) ◽  
pp. 79-88 ◽  
Author(s):  
Said Agamy ◽  
Adul Metwally ◽  
Mohammad Al-Ramady ◽  
Sayed Elaraby
Keyword(s):  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Cooling Water ◽  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Operator Action ◽  
Thermal Hydraulic Analysis

This study describes a RELAP5 computer code for thermal-hydraulic analysis of a typical pressurized water reactor. RELAP5 is used to calculate the thermal hydraulic characteristics of the reactor core and the primary loop under steady-state and hypothetical accidents conditions. New designs of nuclear power plants are directed to increase safety by many methods like reducing the dependence on active parts (such as safety pumps, fans, and diesel generators ) and replacing them with passive features (such as gravity draining of cooling water from tanks, and natural circulation of water and air). In this work, high and medium pressure injection pumps are replaced by passive injection components. Different break sizes in cold leg pipe are simulated to analyze to what degree the plant is safe (without any operator action) by using only these passive components. Also station blackout accident is simulated and the time response of operator action has been discussed.

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 Transformation of Pressurized Water Reactor (AP1000) to Fuzzy Graph

MATEMATIKA ◽  
2018 ◽  
Vol 34 (2) ◽  
pp. 235-244 ◽  
Author(s):  
Azmirul Ashaari ◽  
Tahir Ahmad ◽  
Wan Munirah Wan Mohamad
Keyword(s):  
Nuclear Power ◽  
Ph Value ◽  
Published Data ◽  
Primary System ◽  
Fuzzy Graph ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fission Process ◽  
Water Reactor ◽  
Dynamic Concentration

Pressurized water reactor (PWR) type AP1000 is a third generation of a nuclear power plant. The primary system of PWR using uranium dioxide to generate heat energy via fission process. The process influences temperature, pressure and pH value of water chemistry of the PWR. The aim of this paper is to transform the primary system of PWR using fuzzy autocatalytic set (FACS). In this work, the background of primary system of PWR and the properties of the model are provided. The simulation result, namely dynamic concentration of PWR is verified against published data.

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