scholarly journals Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

2015 ◽  
Author(s):  
J. R. Licht ◽  
A. Bergeron ◽  
B. Dionne ◽  
G. Van den Branden ◽  
S. Kalcheva ◽  
...  
Keyword(s):  
Steady State ◽  
Thermal Hydraulics

Steady-State and Transient Thermal Hydraulics of a Breeder-in-Tube Blanket Design

1986 ◽  
Vol 10 (3P2B) ◽  
pp. 1628-1633 ◽  
Author(s):  
Steven P. Grotz ◽  
Nasr M. Ghoniem
Keyword(s):  
Steady State ◽  
Thermal Hydraulics ◽  
Blanket Design ◽  
Transient Thermal

scholarly journals Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

2021 ◽  
Vol 2021 ◽  
pp. 1-15
Author(s):  
Siyu Lyu ◽  
Daogang Lu ◽  
Danting Sui
Keyword(s):  
Steady State ◽  
Transient Behavior ◽  
Transient Simulation ◽  
Test Facility ◽  
Passive Safety ◽  
Liquid Sodium ◽  
Outlet Temperature ◽  
Thermal Hydraulics ◽  
Safety Margins ◽  
Calculation Module

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

Steady-state thermal-hydraulics feasibility study for the conversion of the BR2 reactor to LEU. Revision 0.

2011 ◽  
Author(s):  
Arne P. Olson ◽  
Benoit Dionne ◽  
John G. Stevens ◽  
S. Kalcheva ◽  
G. Van den Branden ◽  
...  
Keyword(s):  
Steady State ◽  
Feasibility Study ◽  
Thermal Hydraulics

Study on Calculation Difference of the SCWR Core Steady-State Analysis Code

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Lianjie Wang ◽  
Lei Yao ◽  
Ping Yang ◽  
Di Lu ◽  
Wenbo Zhao
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Power Distribution ◽  
Group Structure ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Steady State Analysis ◽  
Energy Group ◽  
The Cross ◽  
State Analysis

Abstract The three-dimensional code system supercritical water-cooled reactor (SCWR) coupled neutronics/thermal-hydraulics analysis (SNTA) code is developed for SCWR core steady-state analysis by coupling neutronics/thermal-hydraulics (N/T). This paper studies the calculation difference between the SNTA code and the standard reactor analysis code (SRAC). By using the impacts exclusive method, it is confirmed that the calculation difference between these two codes is caused by different feedback of the cross section. The cross section data and the energy group structure of the SRAC code differ from the SNTA code, and the density coefficient of reactivity calculated by the SRAC code is higher, which means the feedback of the density and power distribution is bigger and the axial power distribution varies rapidly. The SNTA code with finer energy group structure is suitable for the performance analysis of SCWR core which has strong N/T coupling characteristics.

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