Fuel Assembly Design for Supercritical Water-Cooled Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Feng Linna ◽  
Zhu Fawen
Keyword(s):  
Fuel Assembly ◽  
Supercritical Water ◽  
Nuclear Reactor ◽  
Protective Coatings ◽  
Assembly Design ◽  
Water Reactors ◽  
And Performance ◽  
Materials Used ◽  
Degradation Of Materials ◽  
Supercritical Water Cooled Reactor

The supercritical water-cooled reactor (SWCR) has been selected as one of the most promising reactors for Generation IV nuclear reactors due to its higher thermal efficiency and more simplified structure compared to the state-of-the-art light water reactors (LWRs). However, there are a large number of potential problems that must be addressed, particularly the fuel assembly design of the SCWR. SCWRs are a kind of high-temperature, high-pressure, water-cooled reactor that operates above the thermodynamic critical point of water (374°C, 22.1 MPa). Corrosion and degradation of materials used in supercritical water environments are determined by several environment- and material-dependent factors. In particular, irradiation-induced changes in microstructure and microchemistry are major concerns in a nuclear reactor. Many structural materials including alloys and ceramics have been proposed for use as SCWR components or materials for applying protective coatings in SCWRs. In this paper, the present status of supercritical fuel assembly design at home and abroad is reported. According to the special requirements of supercritical core design, a kind of configuration design of fuel assembly with two-flow core and using SiC as cladding material are proposed. The analysis results have shown that the design basically meets the requirements of fuel assembly design, which has good feasibility and performance.

scholarly journals Optimization of a Fuel Assembly for Supercritical Water-Cooled Reactor CSR1000

2021 ◽  
Vol 9 ◽  
Author(s):  
Fawen Zhu ◽  
Lele Zheng ◽  
Quan-Yao Ren ◽  
Ti Yue ◽  
Hua Pang ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Terminal Velocity ◽  
Structure Design ◽  
Control Rod ◽  
And Performance ◽  
Supercritical Water Cooled Reactor

As one of the Generation IV nuclear reactors, the SCWR (supercritical water-cooled reactor) has high economy and safety margin, good mechanical properties for its high thermal efficiency, and simplified structure design. As the key component of nuclear reactor, the fuel assembly has always been the main issue for the design of the SCWR. The design of the fuel assembly for CSR1000 proposed by the Nuclear Power Institute of China (NPIC) has been optimized and presented in this study, which is composed of four subassemblies welded by four filler strips and guide thimbles arranged close together in the cross-shaped passage. Aiming at improving the hydraulic buffer performance of the cruciform control rod, the scram time and terminal velocity of control rod assembly were calculated to assess the scram performance based on the computational fluid dynamics and dynamic mesh method, and the mechanical property and neutronic performance of assemblies were also investigated. It has been demonstrated that the optimized fuel assembly had good feasibility and performance, which was a promising design for CSR1000.

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.

Various Design Aspects of the Canadian Supercritical Water-Cooled Reactor Core

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Metin Yetisir ◽  
Rui Xu ◽  
Michel Gaudet ◽  
Mohammad Movassat ◽  
Holly Hamilton ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Operating Conditions ◽  
Stress Distributions ◽  
Light Water Reactors ◽  
Core Components ◽  
Water Reactors ◽  
Supercritical Water Cooled Reactor

The Canadian Supercritical Water-Cooled Reactor (SCWR) is a 1200 MW(e) channel-type nuclear reactor. The reactor core includes 336 vertical pressurized fuel channels immersed in a low-pressure heavy water moderator and calandria vessel containment. The supercritical water (SCW) coolant flows into the fuel channels through a common inlet plenum and exits through a common outlet header. One of the main features of the Canadian SCWR concept is the high-pressure (25 MPa) and high-temperature (350°C at the inlet, 625°C at the outlet) operating conditions that result in an estimated thermal efficiency of 48%. This is significantly higher than the thermal efficiency of the present light water reactors, which is about 33%. This paper presents a description of the Canadian SCWR core design concept; various numerical analyses performed to understand the temperature, flow, and stress distributions of various core components; and how the analyses results provided input for improved concept development.

Fuel assembly design study for a reactor with supercritical water

2007 ◽  
Vol 237 (14) ◽  
pp. 1513-1521 ◽  
Author(s):  
J. Hofmeister ◽  
C. Waata ◽  
J. Starflinger ◽  
T. Schulenberg ◽  
E. Laurien
Keyword(s):  
Fuel Assembly ◽  
Supercritical Water ◽  
Design Study ◽  
Assembly Design

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.

Mitigating the Stress Corrosion Cracking of Zircaloy-4 Fuel Sheathing: Siloxane Coatings Revisited

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Graham A. Ferrier ◽  
Mohsen Farahani ◽  
Joseph Metzler ◽  
Paul K. Chan ◽  
Emily C. Corcoran
Keyword(s):  
Wear Resistance ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Reactor ◽  
Corrosion Cracking ◽  
Candu Reactors ◽  
Graphite Coating ◽  
Water Reactors ◽  
And Performance ◽  
Graphite Coatings

For more than 50 years, a thin (3–20 μm) graphite coating has played an important role in limiting the stress corrosion cracking (SCC) of Zircaloy-4 fuel sheathing in CANDU® nuclear reactors. Siloxane coatings, which were examined alongside graphite coatings in the early 1970s, demonstrated even better tolerance against power-ramp-induced SCC and exhibited better wear resistance than graphite coatings. Although siloxane technology developed significantly in the 1980s/1990s, siloxane coatings remain unused in CANDU reactors, because graphite is relatively inexpensive and performs well in-service. However, advanced CANDU designs will accommodate average burnups, exceeding the threshold tolerable by the graphite coating (450  MWh/kgHE). In addition, siloxane coatings may find applicability in pressurized and boiling water reactors, wherein the burnups are inherently larger than those in CANDU reactors. Consequently, a commercially available siloxane coating is evaluated by its present-day chemistry, wear resistance, and performance in hot, stressful, and corrosive environments. After subjecting slotted Zircaloy-4 rings to iodine concentrations exceeding the estimated in-reactor concentration (1  mg/cm3), mechanical deflection tests and scanning electron microscopy (SEM) show that the siloxane coating outperforms the graphite coating in preserving the mechanical integrity of the rings. Furthermore, the baked siloxane coating survived a 50-day exposure to thermal neutron flux ((2.5±0.1)×1011  n/cm2 s) in the SLOWPOKE-2 nuclear reactor at the Royal Military College of Canada.

scholarly journals Investigation of Oxidation Behavior of 304L and 310S Steels with Potential Application in Supercritical Water-cooled Nuclear Reactors

2020 ◽  
Vol 71 (8) ◽  
pp. 98-105
Author(s):  
Florentina Galan ◽  
Marian Catalin Ducu ◽  
Manuela Fulger ◽  
Denis Aurelian Negrea
Keyword(s):  
Supercritical Water ◽  
Nuclear Reactor ◽  
Chromium Content ◽  
Future Research ◽  
Fuel Cladding ◽  
Test Protocol ◽  
Corrosive Media ◽  
Test Parameters ◽  
Surface Oxide Formation ◽  
Materials Used

Selecting proper candidate materials is one key issue for the development of the supercritical water-cooled nuclear reactor (SCWR). Designing or choosing the most fitting materials means better sustainability, economics and safety. As the supercritical water is a very aggressive corrosive media, corrosion becomes one challenging problem for the materials used in the SCWR. This paper involves the corrosion testing of two stainless steels (304L and 310S) and microstructure evaluation of samples after being exposed to supercritical water. The test parameters were set at the temperature of 550oC and the pressure of 25 MPa for up to 63 days. The samples were investigated using gravimetric corrosion test, optical microscopy, scanning electron microscopy coupled with energy dispersive X-ray spectroscopy. Results showed a lower corrosion performance, in terms of weight change and surface oxide formation, for 304L due to its low chromium content. 310S has excellent corrosion resistance because the chromium content is higher. The results obtained will be useful in future research of test protocol and development of alloys that could be used as reactor fuel cladding and other components in the SCWR.

Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Laurence K. H. Leung ◽  
Yanfei Rao ◽  
Krishna Podila
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
Supercritical Pressure ◽  
Heat Transfer Correlation ◽  
Computational Tools ◽  
Experimental Heat ◽  
Experimental Heat Transfer ◽  
Transfer Behavior ◽  
Supercritical Water Cooled Reactor

Experimental data and correlations are not available for the fuel-assembly concept of the Canadian supercritical water-cooled reactor (SCWR). To facilitate the safety analyses, a strategy for developing a heat-transfer correlation has been established for the fuel-assembly concept at supercritical pressure conditions. It is based on an analytical approach using a computational fluid dynamics (CFD) tool and the ASSERT subchannel code to establish the heat transfer in supercritical pressure flow. Prior to the application, the CFD tool was assessed against experimental heat transfer data at the pseudocritical region obtained with bundle subassemblies to identify the appropriate turbulence model for use. Beyond the pseudocritical region, where the normal heat transfer behavior is anticipated, the ASSERT subchannel code also was assessed with appropriate closure relationships. Detailed information on the supporting experiments and the assessment results of the computational tools are presented.

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