Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR

PWR core phenomena can be simulated and predicted more precisely and in more details with high-fidelity neutronics and thermal-hydraulics coupling calculations. An internal coupling between a newly developed high-fidelity neutronics code NECP-X and the sub-channel code SUBSC has been realized. In order to verify the NECP-X/SUBSC coupling system, another high-fidelity neutronics and thermal-hydraulics coupling system OpenMC/SUBSC was developed through external coupling method. Both coupling systems were applied to a simplified PWR 3×3 pin cluster case. The numerical result shows good agreement in both eigenvalue and normalized axial power distribution for a selected pin, demonstrating the success of the internal coupling of NECP-X and SUBSC.

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DEVELOPMENT OF 3D PIN-BY-PIN CORE SOLVER TORTIN AND COUPLING WITH THERMAL-HYDRAULICS

EPJ Web of Conferences ◽

10.1051/epjconf/202124702034 ◽

2021 ◽

Vol 247 ◽

pp. 02034

Author(s):

P. Mala ◽

A. Pautz ◽

H. Ferroukhi ◽

A. Vasiliev

Keyword(s):

Monte Carlo ◽

State Of The Art ◽

The State ◽

Monte Carlo Code ◽

Channel Code ◽

Thermal Hydraulics ◽

Fuel Temperature ◽

Power Profile ◽

Assembly Problem ◽

Good Agreement

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

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Coupled Three Dimensional Neutronics/Thermal-Hydraulics Code STTA for SCWR Core Transient Analysis

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

10.1115/icone22-30078 ◽

2014 ◽

Author(s):

Lianjie Wang ◽

Wenbo Zhao ◽

Ping Yang ◽

Bingde Chen ◽

Dong Yao

Keyword(s):

Transient Analysis ◽

Function Method ◽

Three Dimensional ◽

Supercritical Pressure ◽

Channel Code ◽

Neutron Diffusion ◽

Thermal Hydraulics ◽

The Core ◽

Coupling Approach ◽

Neutron Diffusion Equation

A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN_K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN_K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation.

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COSINE Validation Experiment Plan: 1st Phase

Volume 4: Thermal Hydraulics ◽

10.1115/icone21-15171 ◽

2013 ◽

Author(s):

Lifang Liu ◽

Xiaoliang Fu ◽

Zheng Du ◽

Nan Yu ◽

Yanhua Yang

Keyword(s):

Power Plant ◽

Nuclear Power ◽

System Analysis ◽

Channel Code ◽

Thermal Hydraulics ◽

Plant Design ◽

Code System ◽

International Database ◽

Valuable Data ◽

Validation Experiment

COSINE (COre and System INtegrated Engine for design and analysis) is the software platform for nuclear power plant design and analysis being developed in China, which contains three thermal-hydraulics codes and three physics codes in the first phase of this project. Three thermal-hydraulics codes in COSINE are sub-channel code, system analysis code, and containment analysis code. Separate experiments data and integral experiments data are needed to validate the models and analyze the uncertainty of these codes. Some data are obtained from international database and organizers. At the same time, an experiment plan has been made by SNPSDC to meet the COSINE thermal-hydraulics codes validation requirements. In the 1st phase of this plan, eight separate effects facilities are being constructed to get some valuable data from January 2011 to December 2014, and the test contents has been programmed. The facilities, contents and the schedules of the tests of COSINE validation experiment plan-I are detailed in this paper.

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