TRISO fuel thermal simulations in the LS-VHTR

The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li2BeF4 called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed.

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Reactivity Hold-Down Technique for a Soluble Boron Free PWR Using TRISO Particle Fuel

Volume 5: Fuel Cycle, Radioactive Waste Management and Decommissioning; Reactor Physics and Transport Theory; Nuclear Education, Public Acceptance and Related Issues; Instrumentation and Controls; Fusion Engineering ◽

10.1115/icone21-15275 ◽

2013 ◽

Author(s):

Xiang Dai ◽

Xinrong Cao

Keyword(s):

Nuclear Proliferation ◽

Reactor Core ◽

Fission Products ◽

Fuel Kernel ◽

Average Temperature ◽

Fuel Rod ◽

Triso Fuel ◽

Coated Particle ◽

Graphite Matrix ◽

Triso Particles

TRISO coated particle, developed for HTGR initially, has advantages of nuclear proliferation-resistance and fuel integrity against the release of fission products. In this paper, a 350MWt small sized PWR core design utilizing TRISO fuel concept is presented. TRISO particles are dispersed in graphite matrix to form the fuel compact, and then the fuel compact is clad by Zircaloy-4 cladding to form a fuel rod. The graphite matrix increases thermal conductivity of fuel compact, so that the fuel average temperature would be well below conventional PWRs’. In order to simplify reactor design, operation and maintenance, soluble boron free concept while operation is introduced. The emphasis of the study is put on the reactivity hold-down technique for the 350MWt PWR core. Excess reactivity is suppressed through a combination of Pu-240 adding with Gd2O3 loading. Pu-240 is added into UO2 fuel kernel of some assemblies, and Gd2O3 rods are loaded in other assemblies. The non-fissile plutonium isotope Pu-240 has a considerably high thermal neutron capture cross section compared to U-238, so that the Pu-240 added fuel can greatly suppress excess reactivity over burnup. Besides, reactor core life would be extended by adding proper amount of Pu-240 for its converting into Pu-241 which is a fissile isotope. Combining Pu-240 adding with Gd2O3 loading, the designed core reaches an average core burnup of approximately 58GWD/t, as well as a core life of nearly 6EFPY.

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Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

Advanced Materials Research ◽

10.4028/www.scientific.net/amr.986-987.231 ◽

2014 ◽

Vol 986-987 ◽

pp. 231-234

Author(s):

Jun Teng Liu ◽

Qi Cai ◽

Xia Xin Cao

Keyword(s):

Nuclear Reactor ◽

Natural Circulation ◽

Reactor Core ◽

Saturation Temperature ◽

Heat Removal ◽

Primary Circuit ◽

Station Blackout ◽

Heat Removal System ◽

Removal System ◽

Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

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CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

Volume 9: Student Paper Competition ◽

10.1115/icone26-81626 ◽

2018 ◽

Author(s):

Jing Chen ◽

Dalin Zhang ◽

Suizheng Qiu ◽

Kui Zhang ◽

Mingjun Wang ◽

...

Keyword(s):

Fast Reactor ◽

Power Distribution ◽

Three Dimensional ◽

Reactor Core ◽

Heat Removal ◽

The Core ◽

Computational Fluid Dynamics Cfd ◽

Heat Removal System ◽

China Experimental Fast Reactor ◽

Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

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The Natural Circulation Characteristic Analysis of AP1000 IRWST

Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory ◽

10.1115/icone22-30520 ◽

2014 ◽

Author(s):

Ruojun Xue ◽

Bin Xia ◽

Wang Mingyuan

Keyword(s):

Heat Transfer ◽

Natural Convection ◽

Heat Exchanger ◽

Waste Heat ◽

Natural Circulation ◽

Reactor Core ◽

Heat Removal ◽

Dead Zones ◽

Circulation Characteristic ◽

Heat Removal System

The natural convection passive residual heat removal system is an essential system of AP1000 nuclear power plants, which works in the non-LOCA accident. It can discharge the waste heat in the reactor core timely, and prevent the reactor accident. The In-Containment Refueling Water Storage Tank (IRWST) is one of the most important equipments of passive residual heat removal system, which supplies the hot trap to the system. The purpose of the paper is to research the natural circulation characteristic of fluid in IRWST. In this paper, FLUENT hydrodynamics analysis software was used to simulate the flow and heat-transfer characteristics of natural convection. Temperature field and flow field in different time or locations were compared to analyze the course of natural circulation. The error was evaluated by comparing the average temperature on the outlet of heat exchanger and the design temperature. The result showed that FLUENT could simulate the flow and heat-transfer characteristics of natural convection, and the error of the simulation was acceptable. In conclusion, natural convection heat exchanger of the study can discharge the waste heat in the reactor core timely. Due to the complex structure of the heat exchanger and the IRWST, the “heat-transfer dead zones” and “flow dead zones” are generated in local parts. The presence of dead zones affects the formation of natural convection. With the development of natural circulation, the temperature field and flow field in the IRWST keep stable and the temperate rise rate become slowly.

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Study on a Small Power Reactor With Compact Pressure Vessel and Natural Circulation

Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security ◽

10.1115/icone22-30936 ◽

2014 ◽

Author(s):

Kenya Takiwaki ◽

Shungo Sakurai ◽

Yutaka Takeuchi ◽

Yasushi Yamamoto

Keyword(s):

Power Reactor ◽

Natural Circulation ◽

Reactor Core ◽

Cooling Water ◽

Heat Removal ◽

Small Power ◽

Control Rod ◽

Decay Heat ◽

Heat Removal System ◽

Removal System

There is movement which is developing the small reactor for the small electricity grid in place of a big power reactor which requires the high capital cost. This paper introduces a small power reactor whose purpose is to achieve high economic competitiveness and advanced safety. In order to attain high economic competitiveness, it is designed to be small and simple and uses natural circulation and high pressure. A steam generator is integrated into the reactor pressure vessel (RPV), thus dispensing with a primary system and preventing radiation leakage from the reactor core. The small core is designed to have a high power density (100 MW/m3, almost twice that of a conventional boiling water reactor). The concept of a 300 MWt (100 MWe) core design is established by introducing a boiling heat transfer system. By boiling cooling water, the cooling-water circulating flow quantity in a reactor core is enlarged. By increasing a flow, the minimum critical power ratio is improved, which is an important core characteristic. Furthermore, using a burnable poison (Gd2O3), the excess reactivity of a reactor core is reduced and excess reactivity is controlled only by the control rod. Moreover, the maximum linear power density is improved and the critical power ratio is minimized by optimizing the burnable poison arrangement and the control rod pattern. In order to attain high safety, our small reactor has an advanced decay heat removal system that can cool the core without external support. This decay heat removal system is part of the secondary cooling system and combined with a cooling tower. As a result, the quantity of cooling water stored in the decay heat removal system is reduced, and longtime decay heat removal is possible by small equipment.

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