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 Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

Optimization design for mechanical seals in reactor coolant pumps based on a fluid–solid strong-interaction model

2011 ◽  
Vol 225 (8) ◽  
pp. 1851-1862 ◽  
Author(s):  
C-J Liao ◽  
W-F Huang ◽  
Y-M Wang ◽  
S-F Suo ◽  
X-F Liu
Keyword(s):  
Strong Interaction ◽  
Optimization Design ◽  
Power Plants ◽  
Interaction Model ◽  
Reactor Power ◽  
Mechanical Seals ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
The Operating Mechanism ◽  
And Performance

The study on the mechanism and performance of the mechanical seals in reactor coolant pumps (RCPs) is very important for the safe operations of pressurized water reactor power plants. By exploring the operating mechanism of the first seal of the hydrostatic mechanical seal in RCPs, an analytical fluid–solid strong-interaction model of the seal is proposed in this article. The model holds that the mechanical deformations of the seal assembly are dominated by the deflections of the seal rings, and this idea is demonstrated by the numerical simulation result of a fluid–solid interaction (FSI) model. Using the analytical FSI model, the regularity that the leakage rate of the first seal varies with the differential pressure in a RCP is obtained, and compared with the operational data, which is used to verify the model. Based on the understanding of the reliability of the seal, a dimensionless parameter Λ that acts as an attribute to the reliability is proposed in this article. Using the analytical FSI model and Λ as the optimization algorithm and optimization object, respectively, the optimum designs about the seal faceplateconfigurations are performed. Also, the specific optimization conclusions are given simultaneously.

Transient Thermal Behaviors of SBO Accident for a 200MW OFNP Under Heaving Motion Conditions

2018 ◽  
Author(s):  
Qiqi Yan ◽  
Simin Luo ◽  
Yapei Zhang ◽  
Limin Liu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
System Analysis ◽  
Nuclear Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Thermal Behaviors ◽  
Severe Accidents ◽  
Nuclear Plants ◽  
Station Blackout ◽  
Water Reactors ◽  
Transient Thermal

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Small Modular Reactor and Large Nuclear Reactor Fuel Cost Comparison

2014 ◽  
Author(s):  
Christopher P. Pannier ◽  
Radek Škoda
Keyword(s):  
Nuclear Reactor ◽  
Cost Comparison ◽  
Design Parameters ◽  
Fuel Cost ◽  
Uranium Enrichment ◽  
Pressurized Water ◽  
Nuclear Reactor Fuel ◽  
Water Reactors ◽  
Near Term ◽  
Fuel Costs

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.

How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART

2009 ◽  
Vol 1215 ◽  
Author(s):  
Laurence Luneville ◽  
David Simeone ◽  
Gianguido Baldinozzi ◽  
Dominique Gosset ◽  
yves serruys
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Ion Beam ◽  
Fast Neutrons ◽  
Relaxation Processes ◽  
Pressurized Water ◽  
Water Reactors ◽  
Displacement Per Atom ◽  
The Impact

AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

2021 ◽  
Vol 8 (3B) ◽  
Author(s):  
Thiago Souza Pereira de Brito ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
Wagner Eustáquio de Vasconcelos
Keyword(s):  
Fuzzy Logic ◽  
Pid Controller ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Fuzzy Controller ◽  
Normal Operation ◽  
Primary Circuit ◽  
Level Control ◽  
Pressurized Water ◽  
Safety Margins

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing  pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off).

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