Neutronic and Thermo-Hydraulic Analyses of Water-Cooled Blanket Based on Pressurized/Supercritical Water Conditions for CFETR

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Jie Cheng ◽  
Yingwei Wu ◽  
G. H. Su ◽  
Suizheng Qiu ◽  
Wenxi Tian
Keyword(s):  
Supercritical Water ◽  
Fluid Dynamic ◽  
Safety Margin ◽  
Dominant Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Tokamak Reactor ◽  
Water Conditions ◽  
Neutron Multiplier ◽  
Test Reactor

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on pressurized water and supercritical water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher tritium breeding ratio (TBR) and uniform heat distribution. This blanket adopts the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. At first, the neutronic analysis was performed and based on the typical outboard equatorial blanket. Then, thermal and fluid dynamic analysis of the 3D model was carried out by CFX with fluid–solid coupling approach. It was found that the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water conditions. It indicated that supercritical water blanket had smaller safety margin than pressurized water blanket, but supercritical water blanket would lead to higher outlet temperature, thermal conductivity, and heat exchange efficiency also. In addition, the beryllium fraction was considered as one of the dominant factor, which leading a higher TBR in our simulations.

Neutronics and Thermo-Hydraulic Analysis of Water-Cooled Blanket Based on PWR and SCWR Water Conditions for CFETR

2016 ◽  
Author(s):  
Jie Cheng ◽  
Yingwei Wu ◽  
G. H. Su ◽  
Suizheng Qiu ◽  
Wenxi Tian
Keyword(s):  
Supercritical Water ◽  
Fluid Dynamic ◽  
Safety Margin ◽  
Dominant Factor ◽  
Outlet Temperature ◽  
Pebble Bed ◽  
Pressurized Water ◽  
Tokamak Reactor ◽  
Water Reactor ◽  
Water Conditions

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on PWR (pressurized water reactor) and SCWR (super-critical water reactor) water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher Tritium Breeding Ratio (TBR) and uniform heat distribution. This blanket uses the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. The nuclear heating distribution was obtained from the neutronics calculations by MCNP. The thermal hydraulic behaviors of the first wall (FW), structure material, Li2TiO3 pebble bed and Beryllium pebble bed under normal condition were calculated, respectively. It was found that the temperature on the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water cooling scheme. It indicated that SCWR case had smaller safety margin than PWR case, but SCWR case would lead higher outlet temperature, thermal conductivity and heat exchange efficiency also. In addition, it was found that beryllium was the dominant factor leading a higher TBR. The results would be important to water condition choice for solid blanket in the future.

Irradiation Capabilities at the Halden Reactor and Testing Possibilities Under Supercritical Water Conditions

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Rudi Van Nieuwenhove
Keyword(s):  
Feasibility Study ◽  
Supercritical Water ◽  
External Pressure ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Conditions ◽  
Different Types

Different types of instruments have been developed both for in-pile fuel and materials studies at the Halden Reactor Project. In recent years, several of the standard instruments have been upgraded to be able to tolerate much higher temperatures. In particular, several instruments are now able to operate up to 650°C and 25 MPa, thus in the range suitable for supercritical water (SCW) studies. In addition, a feasibility study for an in-pile SCW loop has been carried out that shows that such a loop can be realized in the Halden reactor, allowing for all the instrumentation possibilities that are presently carried out in pressurized water reactor (PWR) and boiling water reactor (BWR) conditions. Another, and cheaper, alternative is to perform corrosion experiments inside a small capsule in which SCW is maintained by means of gamma heating and external pressure lines. The conceptual designs of the SCW loop and SCW capsule will be highlighted.

Thermo-Hydraulic Analysis of the Optimized Helium Cooled Solid Breeder Blanket for CFETR

2016 ◽  
Author(s):  
Shijie Cui ◽  
Dalin Zhang ◽  
Jie Chang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
...  
Keyword(s):  
Fluid Dynamic ◽  
Critical Conditions ◽  
Inlet Temperature ◽  
Tritium Breeding ◽  
Pebble Bed ◽  
Temperature Distributions ◽  
Neutron Multiplier ◽  
Test Reactor ◽  
Cfd Method ◽  
Solid Breeder Blanket

China Fusion Engineering Test Reactor (CFETR) is under design recently, in which a conceptual structure of the helium-cooled solid breeder blanket is proposed as one of the candidate tritium breeding blankets. In this concept, three radial arranged U-shaped breeding zones are designed and optimized for higher Tritium Breeding Ratio (TBR) and structure simplification. This blanket uses the Li4SiO4 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized typical outboard blanket module are performed by CFD method, where the nuclear heating rate is obtained from the preliminary neutronics calculations. The thermal hydraulic behaviors of the first wall (FW), the temperature distributions of submodule structure material, Li4SiO4 pebble bed and Beryllium pebble bed under normal and critical conditions are calculated, respectively. The results show that the temperature on the blanket module can be effectively cooled below allowable temperature limits of the materials, even if the FW is suffering the maximum surface heat flux, which verified the reasonability of the design of the blanket cooling scheme. In addition, several parametric sensitivity studies are conducted to investigate the influences of main parameters (e.g. coolant mass flow rate, inlet temperature, pebble bed thermal conductivity and fusion power) on the temperature distributions of the blanket components.

Conceptual Design of the Water Cooled Breeder Blanket for Both Phases of CFETR

2018 ◽  
Author(s):  
Songlin Liu ◽  
Xuebin Ma ◽  
Kecheng Jiang ◽  
Min Li ◽  
Xiaokang Zhang
Keyword(s):  
Phase I ◽  
Phase Ii ◽  
Outlet Temperature ◽  
Allowable Limit ◽  
Pressurized Water ◽  
Design Work ◽  
Fusion Power ◽  
Blanket Design ◽  
Neutron Multiplier ◽  
Two Phases

The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on fusion power Pfusion = 200 MW, fusion power gain Qplasma = 1 – 5, tritium breeding ratio TBR>1.0, neutron DPA requirement ∼10 dpa. Phase II emphasizes DEMO validation, which means Qplasma > 10, Pfusion > 1 GW, e.g. 1.5 GW. It is required that one blanket design can cover the operation of both phases of CFETR from the viewpoint of saving construction cost and reducing waste inventory. However, fusion power in Phase-II is 4–6.5 times larger than those in Phase-I, and this also causes the great challenge facing the thermal-hydraulics design of the blanket. A new version of water cooled ceramic breeder (WCCB) blanket for both phases is proposed for CFETR, based on a trade-off considering on TBR, release tritium temperature in breeder zone, and removal heat capability of coolant. This design continues to employ the mixed breeder of Li2TiO3 and Be12Ti as tritium breeder and primary neutron multiplier, and a few Be as supplement of multiplying neutrons, Reduced Activation Ferritic/Martensitic steel as structural material, tungsten as plasma facing material. Pressurized water of 15.5 MPa is chosen as coolant with 285 °C inlet/325 °C outlet temperature. Main design change is that it employs two independent coolant systems in the blanket cooling components. For Phase I, one coolant system is only used and hoped to improve the breeder zone temperature higher than tritium release temperature. For Phase II, all of two coolant systems are put into using to ensure the material temperature less than the allowable limit. In this paper, the WCCB blanket design work is presented and its feasibility is investigated from the aspect of neutronics and thermo-hydraulics.

Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle

2009 ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyy ◽  
A. Eu. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Annular Channel ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter

The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

Thermal Aspects of Using Uranium Mononitride Fuel in a SuperCritical Water-Cooled Reactor at Maximum Heat Flux Conditions

2010 ◽  
Author(s):  
Ashley Milner ◽  
Caleb Pascoe ◽  
Hemal Patel ◽  
Wargha Peiman ◽  
Graham Richards ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Thermal Efficiency ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Maximum Heat ◽  
Light Water ◽  
Maximum Heat Flux ◽  
Centerline Temperature ◽  
Uranium Mononitride

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.

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.

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