Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Previously reported computations are extended to analyze a Bree plate subjected to steady mechanical load and cyclic thermal downshocks on both surfaces. It is shown that two critical plate thicknesses can be identified, one below which thermal cycling has no effect, and one above which the full effects of rapid transient thermal loading are observed. The significant effects which material properties, such as the degree of strain hardening present, can have on the plate behavior are also illustrated. These findings are relevant to design of fast reactor components which experience loadings of this type.

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Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II

Volume 2A: Thermal Hydraulics ◽

10.1115/icone22-30485 ◽

2014 ◽

Author(s):

Yue Nina ◽

Ma Zaiyong ◽

Hu Benxue ◽

Suizheng Qiu ◽

Guanghui Su

Keyword(s):

Fast Reactor ◽

Channel Model ◽

Flow Characteristics ◽

Primary Loop ◽

Reactor Kinetics ◽

Multiple Channel ◽

Benchmark Analysis ◽

The Core ◽

Intermediate Heat Exchanger ◽

Transient Thermal

In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.

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Development of a transient thermal-hydraulic code for analysis of China Demonstration Fast Reactor

Annals of Nuclear Energy ◽

10.1016/j.anucene.2012.12.022 ◽

2013 ◽

Vol 55 ◽

pp. 302-311 ◽

Cited By ~ 7

Author(s):

B.X. Hu ◽

Y.W. Wu ◽

W.X. Tian ◽

G.H. Su ◽

S.Z. Qiu

Keyword(s):

Fast Reactor ◽

Transient Thermal

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The Accumulation of Damage in Plates Subjected to Steady Mechanical Load and Cyclic Thermal Downshocks on One Surface

Journal of Pressure Vessel Technology ◽

10.1115/1.3264210 ◽

1982 ◽

Vol 104 (3) ◽

pp. 232-238 ◽

Cited By ~ 1

Author(s):

J. Phillips

Keyword(s):

Fatigue Damage ◽

Fast Reactor ◽

Damage Accumulation ◽

Mechanical Load ◽

Computer Calculations ◽

Accumulation Of Damage ◽

Transient Thermal ◽

Biot Numbers ◽

Damage Summation ◽

Dwell Period

Computer calculations, incorporating analysis of elastic, plastic and creep effects, are performed on a Bree plate subjected to steady mechanical load and cyclic thermal downshocks on one surface. Particular attention is given to the accumulation of creep and fatigue damage which is calculated from the stress-strain distributions using linear damage summation rules. The computations examine the dependence of the damage accumulation on such factors as the applied loads, the degree of hardening in the material, the length of the creep dwell period and the severity of the transient thermal loadings as measured by the Fourier and Biot numbers. This indicates which considerations are likely to be important in the design of fast reactor components which experience loadings of this type.

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Research and development of a transient thermal–hydraulic code for system safety analysis of sodium cooled fast reactor

Annals of Nuclear Energy ◽

10.1016/j.anucene.2020.107841 ◽

2021 ◽

Vol 152 ◽

pp. 107841

Author(s):

Yang Yu ◽

Dong Liu ◽

Xiaoming Song ◽

Zhongchun Li ◽

Fengchen Guo ◽

...

Keyword(s):

Research And Development ◽

Fast Reactor ◽

Safety Analysis ◽

System Safety ◽

Transient Thermal

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