scholarly journals Thermal hydraulic analysis of a PWR loaded with annular fuel rods

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Wallen Ferreira De Souza ◽  
Maria Auxiliadora Fortini Veloso ◽  
Antonella Lombardi Costa ◽  
Clubia Pereira
Keyword(s):  
High Performance ◽  
Nuclear Energy ◽  
Nuclear Reactor ◽  
Final Report ◽  
Safety Margins ◽  
Hydraulic Conditions ◽  
Fuel Design ◽  
Energy Center ◽  
Thermal Hydraulic Analysis ◽  
Reference Document

In 2006, the final report of the MIT Center for Advanced Nuclear Energy Center the project entitled High Performance Fuel Design for Next Generation PWR’s presented the proposal of an internal and external cooled ring fuel with the objective of increasing the power density of a PWR reactor without compromising the safety margins of the installation. The thermal hydraulic conditions were calculated with the aid of the VIPRE subchannel code, which is a widely used tool in the analysis of nuclear reactor cores. STHIRP-1 is a subchannel code that has been developed at the Departamento de Engenharia Nuclear /UFMG. In order to evaluate the capacity of the STHIRP-1 program, mainly in relation to the thermal model, the new fuel concept was analyzed. The results were compared with those performed with the VIPRE code presented in the reference document.

scholarly journals Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

2008 ◽  
Vol 2008 ◽  
pp. 1-9 ◽  
Author(s):  
Enrico Zio ◽  
Francesco Di Maio
Keyword(s):  
Confidence Interval ◽  
Order Statistics ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Fuel Cladding ◽  
Complete Group ◽  
Bootstrap Technique ◽  
Safety Margins ◽  
Group Distribution

In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Experimental Investigation on Dependence of Decontamination Factor on Aerosol Number Concentration in Pool Scrubbing Under Normal Temperature and Pressure

2018 ◽  
Author(s):  
HaoMin Sun ◽  
Shinichi Machida ◽  
Yasuteru Sibamoto ◽  
Yuria Okagaki ◽  
Taisuke Yonomoto
Keyword(s):  
High Performance ◽  
Nuclear Reactor ◽  
Experimental System ◽  
Decontamination Factor ◽  
Number Concentration ◽  
Severe Accident ◽  
Measurement Instruments ◽  
Aerosol Number Concentration ◽  
Real Phenomenon ◽  
Validation Tests

During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption. In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.

ACR-1000™ Fuel Design Verification

2010 ◽  
Author(s):  
David J. Wren ◽  
Patrick Reid ◽  
Len L. Wright
Keyword(s):  
High Performance ◽  
Reactor Core ◽  
Cold Test ◽  
Performance Criteria ◽  
Extensive Experience ◽  
Fuel Bundle ◽  
Test Loop ◽  
Fuel Design ◽  
And Performance ◽  
The Impact

The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.

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