Coupled Neutronics/Thermal-Hydraulics Analysis of PT–SCWR With 54-Element Bundle Design

2013 ◽  
Author(s):  
Jia Feng ◽  
Jianqiang Shan ◽  
Bo Zhang
Keyword(s):  
Surface Temperature ◽  
Power Distribution ◽  
Thermal Hydraulics ◽  
Water Coolant ◽  
Uranium Enrichment ◽  
Reference Case ◽  
Light Water ◽  
Lower Maximum ◽  
Tube Type ◽  
Water Moderator

The radial power distribution is uneven in the Pressure Tube type SCWR bundle because of the higher fuel enrichment and the separation of light water coolant and heavy water moderator. The uneven profile will usually result in high cladding temperature. In order to optimize bundle design, the analysis of coupled neutronics/thermal-hydraulics was carried out. The neutronics code WIMS-AECL and the subchannel code ATHAS are selected to conduct the coupled neutronics/thermal-hydraulics analysis for the PT-SCWR 54-elements bundle design. The analysis of the reference case showed that the maximum cladding surface temperature of the bundles at both Beginning of the Cycle (BOC) and End of the Cycle (EOC) exceed the design limits. The optimization was carried out through (1) adopting staggered radial uranium enrichment profile and (2) optimizing the pitch-circle diameters of element rings. The results showed that the optimized bundle can achieve lower maximum cladding surface temperature, which is well below the limiting criteria, and meet the criteria of neutronics demands.

ACR-1000 Passive Features

2006 ◽  
Author(s):  
B. Lekakh ◽  
K. Hau ◽  
S. Ford
Keyword(s):  
Heavy Water ◽  
Reactor Design ◽  
Water Coolant ◽  
Design Elements ◽  
Light Water ◽  
Type Reactor ◽  
Safety Margins ◽  
Passive Design ◽  
Tube Type ◽  
Water Moderator

The Advanced CANDU Reactor™ (ACR™) is a Generation III+ pressure tube type reactor using light water coolant and heavy water moderator. The ACR-1000 reactor design is an evolutionary extension of the proven CANDU reactor design. The ACR-1000 incorporates multiple and diverse passive systems for accident mitigation. Where necessary, one or more features that are passive in nature have been included for mitigation of any postulated accident event. This paper describes how the use of passive design elements complements active features enhances reliability and improves safety margins.

Assessment of the Neutronic Characteristics of DCA Lattices

2005 ◽  
Vol 277-279 ◽  
pp. 747-752
Author(s):  
B.J. Min ◽  
W.Y. Kim
Keyword(s):  
Heavy Water ◽  
Core Structure ◽  
Thermal Reactor ◽  
Multiplication Factor ◽  
Research Facility ◽  
Light Water ◽  
The Core ◽  
Tube Type ◽  
Neutronic Characteristics ◽  
Water Moderator

An investigation on the effect of the neutronic behavior on the lattice for a Deuterium Critical Assembly (DCA) in JNC (Japan Nuclear Cycle Development Institute) has been performed. The DCA, the heavy water moderated and light water cooled pressure-tube type research facility, was designed not only for the core physics research, but also for the development of the core-related technology for the Advanced Thermal Reactor (ATR). The core structure of the ATR is highly heterogeneous and it is separated from the heavy water moderator by a calandria tube. Therefore, the neutron behavior is quite complicated and sensitive to a change of the core structure. In this study, the assessment of the core physics characteristics such as the multiplication factor and the void coefficient for the DCA was conducted using the WIMS-D5 code and the results were compared with those of both the experimental data and WIMS-AECL.

Fully Coupled Simulation of Oxygen and Heat Diffusion for (U,Pu)O2 Fuel in Both Fast-Breeder Reactor and Light-Water Reactor

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Wenzhong Zhou ◽  
Rong Liu
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Heat Diffusion ◽  
Light Water Reactor ◽  
Fast Breeder Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Ratio Change ◽  
Overall Performance ◽  
Fully Coupled

Oxygen redistribution with a high-temperature gradient is an important fuel performance concern in fast-breeder reactor (FBR) and light-water reactor (LWR) (U,Pu)O2 fuel under irradiation, and affects fuels properties, power distribution, and fuel overall performance. This paper studies the burnup dependent oxygen and heat diffusion behavior in a fully coupled way within (U,Pu)O2 FBR and LWR fuels. The temperature change shows relatively larger impact on oxygen to metal (O/M) ratio redistribution rather than O/M ratio change on temperature, whereas O/M ratio redistributions show different trends for FBR and LWR fuels due to their different deviations from the stoichiometry of oxygen under high-temperature environments.

Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics

2010 ◽  
Author(s):  
Zhenyang Li ◽  
Tao Zhou ◽  
Canhui Sun ◽  
Xiaozhuang Liu
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Power Distribution ◽  
Neutron Cross Section ◽  
Physical Characteristics ◽  
Supercritical Pressure ◽  
Light Water ◽  
Assembly Design ◽  
Temperature Distributions ◽  
Parallel Channels

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

Preliminary Verification of the High-Fidelity Neutronics and Thermal-Hydraulics Coupling System NECP-X/SUBSC

2017 ◽  
Author(s):  
Jun Chen ◽  
Liangzhi Cao ◽  
Zhouyu Liu ◽  
Hongchun Wu ◽  
Yijun Zhang
Keyword(s):  
Power Distribution ◽  
Coupling Method ◽  
High Fidelity ◽  
Channel Code ◽  
Thermal Hydraulics ◽  
Numerical Result ◽  
Coupling System ◽  
External Coupling ◽  
Internal Coupling ◽  
Good Agreement

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.

scholarly journals Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Baowen Yang ◽  
Jianqiang Shan ◽  
Junli Gou ◽  
Hui Zhang ◽  
Aiguo Liu ◽  
...  
Keyword(s):  
Power Distribution ◽  
Fundamental Difference ◽  
Heat Fluxes ◽  
Correction Factors ◽  
Light Water Reactors ◽  
Rod Bundles ◽  
Light Water ◽  
Driving Mechanisms ◽  
Rod Bundle ◽  
Water Reactors

Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD). The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

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