Nuclear reactor instrumentation. Pressurized water reactor (PWR) of VVER design. Monitoring adequate cooling within the core during cold shutdown

2001 ◽  
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core

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 Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

A Systematic Review of PCTRAN-Based Pressurized Water Reactor Transient Analysis

2021 ◽  
Author(s):  
Suubi Racheal ◽  
Yongkuo Liu ◽  
Miyombo Ernest ◽  
Abiodun Ayodeji
Keyword(s):  
Nuclear Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Research Education ◽  
User Expertise ◽  
Scale Down ◽  
Water Reactors ◽  
The Impact ◽  
And Training

Abstract The impact of nuclear accidents has been a topic of debate since the construction of the first nuclear reactor, and still stands as a key issue of public concern. Several codes and simulators have been used to study the transient progression in pressurized water reactors, and to evaluate the technical measures adopted to scale down the risk of accidents. However, some of these codes are not suitable for multipurpose research and training as they require significant user expertise, leading to analysis uncertainties largely from the code user effect. This paper presents a bird-eye view of one of the most widely used nuclear reactor transient analyzer — the Personal Computer Transient Analyzer (PCTRAN). This paper discusses the comparative advantages of the simulator from the users’ perspective, with specific attention to its utilization both for research and training. The paper also demonstrates the ease of usage by simulating common transient in a pressurized water reactor. Finally, observations and possible improvements to the code to increase its usability in research, education and training are discussed. This work aims to evaluate the robustness of the simulator towards better utilization for research and training, especially in nuclear newcomer countries.

The CANDLE burnup strategy applied to small modular pressurized water reactor loading with fully ceramic microencapsulated fuel

Kerntechnik ◽  
2022 ◽  
Vol 0 (0) ◽  
Author(s):  
Jinfeng Huang ◽  
Jiaming Jiang
Keyword(s):  
Equilibrium State ◽  
Power Plants ◽  
Reactor Core ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Core Height ◽  
Optimization Schemes

Abstract For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Numerical Simulation of the Transient Phase Change Heat Transfer in a Pressurized Water Reactor Core During a Loss-of-Coolant Accident

2011 ◽  
Author(s):  
Soo W. Jo ◽  
Yong K. Lee ◽  
Jong C. Jo
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Transient Heat Transfer ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.

Inherent Boron Dilution Safety Issue in the French Pressurized Water Reactor: CFD Approach

2006 ◽  
Author(s):  
Estelle Graffard ◽  
Fre´de´ric Goux
Keyword(s):  
Boron Concentration ◽  
Safety Issue ◽  
Natural Circulation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Borated Water ◽  
Steam Generator Tubes ◽  
Boron Dilution

Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed.

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