Thermal‐hydraulic analysis of UO 2 and MOX fuel considering different cladding materials at various burnup levels in pressurized water reactor

Heat Transfer ◽  
2021 ◽  
Author(s):  
Farhana Islam Farha ◽  
Md. Hossain Sahadath ◽  
Sadek Hossain Nishat
Keyword(s):  
Pressurized Water Reactor ◽  
Hydraulic Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
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 Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

Research on the Loading Pattern of Reactor Core With MOX Fuel in the First Cycle

2012 ◽  
Author(s):  
Tianqi Zhang ◽  
Shihe Yu ◽  
Xinrong Cao
Keyword(s):  
Cross Section ◽  
Reactor Core ◽  
Fuel Utilization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loading Pattern ◽  
Mox Fuel ◽  
Initial Cycle ◽  
Calculation Code

In order to research the performance of Pressurized Water Reactor (PWR) with 1/3 MOX fuel in the initial cycle, this paper serves Qinshan II reactor core as the reference core to design suitable MOX assemblies and study relevant core properties. The analyses documented within use assembly cross section calculation code CASMO-4 and core calculation code SIMULATE-3 studied by Studsvik. The purpose of this paper is to demonstrate that the Qinshan II reactor is capable of complying with the requirement for MOX fuel utilization without significant changes to the design of the plant. Several impacts on key physics parameters and safety analysis assumptions, introduced by MOX, are discussing in the paper.

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.

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