Validation of plant dynamics analysis code for fast reactor core thermal hydraulics under natural circulation conditions

Stability analyses are carried out for a natural circulation lead-cooled fast reactor which is at its design stage. Because of its natural circulation feature, flow stability as well as nuclear-coupled thermal hydraulics stability is of some concern. Modal Expansion Method (MEM) and Reduced Order Model (ROM) are used for neutronics and thermal hydraulics part, respectively. These two separate models are coupled through feedback effects. Different simulation scenarios on the separate and coupled models are reported.

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Transient experiments on fast reactor core thermal-hydraulics and its numerical analysis

Nuclear Engineering and Design ◽

10.1016/s0029-5493(99)00324-6 ◽

2000 ◽

Vol 200 (1-2) ◽

pp. 157-175 ◽

Cited By ~ 17

Author(s):

M. Nishimura ◽

H. Kamide ◽

K. Hayashi ◽

K. Momoi

Keyword(s):

Numerical Analysis ◽

Fast Reactor ◽

Reactor Core ◽

Thermal Hydraulics ◽

Transient Experiments

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Integrated Multi-Modular Fast Reactor Concept Design

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54972 ◽

2012 ◽

Cited By ~ 2

Author(s):

Vishal Patel ◽

Pavel Tsvetkov

Keyword(s):

Fast Reactor ◽

Reactor Core ◽

Negative Temperature ◽

Thermal Hydraulics ◽

Concept Design ◽

Fuel Temperature ◽

Feasibility Assessment ◽

Balance Of Plant ◽

Critical Reactor ◽

Economics Analysis

The Integrated Multi-Modular Fast reactor is a pre-conceptual small modular fast reactor design consisting of 7 self-consist subcritical modules, each utilizing a BeO-MOX concept fuel with complete supercritical CO2 brayton cycle turbo-machinery. The subcritical modules, when brought into proximity of one another, form a complete critical reactor core. The feasibility of the reactor is assessed on a full core level, which includes a neutronics, thermal hydraulics, balance of plant, economics, and economics analysis. It has been shown that a critical configuration lasting for 14 years at 10 MWth can be achieved. A hot channel thermal hydraulics and safety analysis shows that the reactor can operate within safety limits with negative temperature coefficients of reactivity as well as stay within fuel temperature limits. A plant thermal efficiency of 36% was achieved and there is room for optimization to achieve higher efficiencies. An economical feasibility assessment shows that the reactor can be economical based on an economy of serial production argument. The analysis also leads to a licensing discussion.

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CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Science and Technology of Nuclear Installations ◽

10.1155/2016/9612120 ◽

2016 ◽

Vol 2016 ◽

pp. 1-12 ◽

Cited By ~ 1

Author(s):

Pengcheng Zhao ◽

Kangli Shi ◽

Shuzhou Li ◽

Jingchao Feng ◽

Hongli Chen

Keyword(s):

Steady State ◽

Fast Reactor ◽

Cooling System ◽

Natural Circulation ◽

Reactor Core ◽

Transient State ◽

Fuel Pin ◽

Gen Iv ◽

Steady State Simulation ◽

Accident Conditions

Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

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