An improved core thermal-hydraulic model for coastdown transient in pressurized water reactor

2017 ◽  
Vol 232 (4) ◽  
pp. 416-424
Author(s):  
Idrees Ahmad ◽  
Saad Arshad ◽  
Sajjad Tahir ◽  
Qaisar Nadeem ◽  
Abdus Samee
Keyword(s):  
Time Constant ◽  
Thermal Model ◽  
Nuclear Reactor ◽  
Transient Behavior ◽  
Nucleate Boiling ◽  
Hydraulic Model ◽  
Half Time ◽  
Accurate Analysis ◽  
Pressurized Water ◽  
Reactor Coolant

The safety of the nuclear reactor revolves around the accurate analysis of rapid flow transient for the design and manufacturing of reactor coolant pumps. In this article, the coastdown transient initiated by the loss of offsite power is simulated. In this case, the pumps are operated by the inertia of the flywheel, therefore, the reliable operation of reactor coolant pumps is the key to the safety of the nuclear reactor. A new hydraulic, as well as the thermal model, is developed for simulating various core parameters during the coastdown period. The present hydraulic model accounts for both the pump half-time and the loop half-time, which is used to increase the accuracy of predicted results over a longer period of time. The results predicted by the hydraulic model are incorporated into a thermal model, which also includes the decay heat following the reactor shutdown. This new model depends upon the core time constant, loop time constant, pump half-time, and hydraulic constant coefficient. The predicted results of flow rate, pressure, temperature, and departure from nucleate boiling ratio are compared with the experimental data and have found good agreement between the two cases. Finally, the departure from nucleate boiling ratio shows that the transient behavior of the reactor is moving toward safety.

scholarly journals Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

2012 ◽  
Vol 66 (3) ◽  
pp. 291-299 ◽  
Author(s):  
Grégory Lefèvre ◽  
Ljiljana Zivkovic ◽  
Anne Jaubertie
Keyword(s):  
Nuclear Reactor ◽  
Primary Circuit ◽  
Particle Removal ◽  
Mechanical Seals ◽  
Solution Ph ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Key Factor ◽  
Water Reactors ◽  
Chemical Conditions

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition) in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek) theory and used as such to interpret this industrial phenomenon.

scholarly journals Approximate model for evaluation of thermal-hydraulic transients associated with rapid power increase in research nuclear reactor

2020 ◽  
Vol 35 (4) ◽  
pp. 310-315
Author(s):  
Maciej Lipka ◽  
Gawel Madejowski ◽  
Rafal Prokopowicz ◽  
Krzysztof Pytelt
Keyword(s):  
Nuclear Fuel ◽  
Time Constant ◽  
Wall Temperature ◽  
Nuclear Reactor ◽  
Hot Spot ◽  
Nucleate Boiling ◽  
Approximate Model ◽  
Fuel Element Cladding ◽  
Proposed Model ◽  
Thermal Calculations

A simple model, for the estimation of changes in the nuclear fuel element cladding temperature as well as the conditions of the formation of the onset of nucleate boiling, is proposed. The results of this estimation are sufficient to assess the effect of the transient with the peak of the reactor's power on the device's safety, without the need for time-consuming thermal calculations. The basic parameters determined using the proposed model are the maximum wall temperature of the device in a hot spot, the time constant of the wall temperature change, and the course of changes in the onset of nucleate boiling ratio in time. The model may be used for investigating the thermal behavior of thin heat-generating and water-cooled elements (such as fuel elements or uranium irradiation targets) during rapid power rise. The results of the temperature estimation with the proposed model were tested considering the hot spot in the MR-6 type nuclear fuel. The SN code with coupled kinetics and thermal-hydraulics, developed in the MARIA reactor was used to validate the results. The maximum cladding temperature, the thermal time constant and the onset of nucleate boiling ratio parameter simulated by the SN code and the proposed scheme appeared to be very consistent.

CFD ANALYSIS OF THE WATER/ALUMINA NANOFLUID AS A NUCLEAR REACTOR COOLANT

2020 ◽  
Author(s):  
Luiz Carlos Aldeia Machado ◽  
Carolina da Silva Bourdot Dutra ◽  
Gabriel Caetano Gomes Ribeiro da Silva
Keyword(s):  
Nuclear Reactor ◽  
Cfd Analysis ◽  
Reactor Coolant ◽  
Alumina Nanofluid

Numerical and Experimental Research on the Fluid-Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

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