Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

2002 ◽  
Author(s):  
Woon-Shing Yeung ◽  
Ramu K. Sundaram
Keyword(s):  
Coming Out ◽  
Computer Code ◽  
Nitrogen Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Concentration Difference ◽  
Dissolved Nitrogen ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Release Model

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

scholarly journals Loss of Coolant Accident in Pressurized Water Reactor. Prediction of a 6-inch Cold Leg Break with Relap5 and Cathare 2

2016 ◽  
Vol 157 ◽  
pp. 333-340 ◽  
Author(s):  
Michał Pawluczyk ◽  
Piotr Mazgaj ◽  
Sebastian Gurgacz ◽  
Michał Gatkowski ◽  
Piotr Darnowski
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Improvement of the CATHARE 3D Code Prediction on PIERO Transient

2008 ◽  
Author(s):  
Tony Glantz ◽  
Roberto Freitas
Keyword(s):  
Direct Application ◽  
Full Scale ◽  
The Other ◽  
Second Step ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Lower Head ◽  
Scale Modelling

The PIERO experiment has been carried out to study phase’s separation in the lower plenum and the downcomer of a Pressurized Water Reactor (PWR) during the end of the depressurization of a large break loss of coolant accident (LB-LOCA). This experiment has been used for the validation and assessment of the 3D module of CATHARE code [1] but the results are not good because of an overestimation of the liquid entrainment in the lower plenum in one hand and the use of a coarsed meshing for modelling the PIERO experiment in the other hand. Two ways of improvement are possible: the first one and the most complicated is to introduce a stratification model in the 3D module of CATHARE. The other one is a possibility to use a refining meshing in order to simulate PIERO experiment. This second way has been performed and the computations results are greatly improve. Nevertheless, PIERO experiment is not on a reactor scale and a direct application of the meshing recommendations made on PIERO is impossible to translate directly on the reactor case. So, the strategy of validation applied to the reactor case consisted in reproducing a PIERO transient with a full scale lower plenum in a first step. In a second step, a converged meshing for the full scale modelling has been determined. In a last step, results obtained with this kind of modelling have been validated via two correlations developed by Wallis and al., that define boundaries conditions between which the water level remaining in the lower head is allowed to vary. This strategy of validation led to model the reactor’s lower plenum with the more axial meshes in order to have good results.

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.

Sign in / Sign up

Export Citation Format

Share Document