Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor

2014 ◽  
Vol 63 ◽  
pp. 587-593 ◽  
Author(s):  
L. Blomeley ◽  
J. Pencer ◽  
B. Hyland ◽  
F.P. Adams
Keyword(s):  
Supercritical Water ◽  
Nuclear Data ◽  
Supercritical Water Cooled Reactor ◽  
Data Sensitivity

scholarly journals Investigation of Corrosion Resistance of Alloys with Potential Application in Supercritical Water-cooled Nuclear Reactors

2019 ◽  
Vol 63 (2) ◽  
pp. 328-332 ◽  
Author(s):  
Ákos Horváth ◽  
Attila R. Imre ◽  
György Jákli
Keyword(s):  
Corrosion Resistance ◽  
Supercritical Water ◽  
Power Plants ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactors ◽  
Reactor Types ◽  
Materials Used ◽  
Supercritical Water Cooled Reactor ◽  
Nuclear Applications

The Supercritical Water Cooled Reactor (SCWR) is one of the Generation IV reactor types, which has improved safety and economics, compared to the present fleet of pressurized water reactors. For nuclear applications, most of the traditional materials used for power plants are not applicable, therefore new types of materials have to be developed. For this purpose corrosion tests were designed and performed in a supercritical pressure autoclave in order to get data for the design of an in-pile high temperature and high-pressure corrosion loop. Here, we are presenting some results, related to corrosion resistance of some potential structural and fuel cladding materials.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Rate Theory for Dislocation Loops Evolution in AL-6XN Austenitic Stainless Steel under Proton Irradiation

2018 ◽  
Vol 913 ◽  
pp. 237-246 ◽  
Author(s):  
Yan Xia Yu ◽  
Li Ping Guo ◽  
Zheng Yu Shen ◽  
Yun Xiang Long ◽  
Zhong Cheng Zheng ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Supercritical Water ◽  
Proton Irradiation ◽  
Theory Model ◽  
Average Density ◽  
Rate Theory ◽  
Dislocation Loops ◽  
Average Size ◽  
Supercritical Water Cooled Reactor

The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.

Flow instability analysis of supercritical water-cooled reactor CSR1000 based on frequency domain

2012 ◽  
Vol 49 ◽  
pp. 70-80 ◽  
Author(s):  
Xiaoyan Tian ◽  
Wenxi Tian ◽  
Dahuan Zhu ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Frequency Domain ◽  
Supercritical Water ◽  
Flow Instability ◽  
Instability Analysis ◽  
Supercritical Water Cooled Reactor

European supercritical water cooled reactor

2011 ◽  
Vol 241 (9) ◽  
pp. 3505-3513 ◽  
Author(s):  
T. Schulenberg ◽  
J. Starflinger ◽  
P. Marsault ◽  
D. Bittermann ◽  
C. Maráczy ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Supercritical Water Cooled Reactor

Improving on Assembly of SCWR Using MOX Fuel

2011 ◽  
Vol 347-353 ◽  
pp. 1633-1636 ◽  
Author(s):  
Can Hui Sun ◽  
Tao Zhou ◽  
Zhou Sen Hou ◽  
Meng Ying Liu ◽  
Feng Luo
Keyword(s):  
Supercritical Water ◽  
Power Distribution ◽  
Security Issue ◽  
Original Design ◽  
Fuel Rod ◽  
Simple Power ◽  
Mox Fuel ◽  
Higher Temperature ◽  
The One ◽  
Supercritical Water Cooled Reactor

A calculation is made for certain Supercritical Water Cooled Reactor (SCWR) using UO2 fuel and MOX fuel respectively. The results indicate that MOX fuel has a simple power distribution with UO2 fuel, but there is a larger power uneven factor when using MOX fuel, and using MOX fuel including weapon grade Pu has larger power uneven factor than using MOX fuel including reactor grade Pu. However, in the case of same power distribution, the fuel rod using MOX fuel has a higher temperature than the one using UO2 fuel. Therefore with the more uneven power distribution, the fuel in SCWR using MOX fuel has a higher temperature. This will result in a big security issue when using MOX fuel in original design of SCWR. Through analyzing the result of power distribution, an improved assembly of SCWR is presented. It can reduce the power uneven factor and increase the security of fuel rod using the improved assembly of SCWR.

Pressure Drop Analysis of a Pressure-Channel Type SuperCritical Water-Cooled Reactor

2013 ◽  
Author(s):  
A. Dragunov ◽  
W. Peiman
Keyword(s):  
Pressure Drop ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Hydraulic Analysis ◽  
Pressure Drops ◽  
One Dimensional ◽  
Supercritical Water Cooled Reactor ◽  
Thermal Hydraulic Analysis ◽  
Pressure Channel ◽  
Channel Type

Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-cooled Reactor (SCWR) and to calculate the temperature profile of the sheath and the fuel bundles. One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions.

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