scholarly journals AN IMPROVED ENERGY DEPOSITION MODEL IN MPACT AND EXPLICIT HEAT GENERATION COUPLING WITH CTF

2021 ◽  
Vol 247 ◽  
pp. 02033
Author(s):  
Yuxuan Liu ◽  
Robert Salko ◽  
Kang Seog Kim ◽  
Xinyan Wang ◽  
Matthew Kabelitz ◽  
...  
Keyword(s):  
Energy Release ◽  
Heat Generation ◽  
Power Distribution ◽  
Energy Deposition ◽  
Reactor Core ◽  
Fission Neutron ◽  
Guide Tube ◽  
Neutron Fission ◽  
Deposition Model ◽  
The Time Domain

The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation of delayed neutron precursors. To allow the explicit heat generation coupling, the interfaces between MPACT and CTF were updated to transfer separate heat sources for different material regions (fuel, clad, moderator and guide tube). The results show that the distributions of the energy deposition between MPACT and MCNP agree very well for various 2-D assembly and quarter-core problems without TH feedback. The MPACT/CTF coupled calculation for the hot full power quarter-core case exhibited a reduced peak pin power by 2.3% and a reduced peak fuel centerline temperature by 17 K when using the explicit energy deposition and heat transfer. The new model also shows a maximum 100 pcm keff effect on assembly depletion problems and an increased overall energy release by 7% in a PWR reactivity-initiated accident (RIA) problem.

Simulation of pulsed accidental energy release in a reactor core

Atomic Energy ◽  
1994 ◽  
Vol 77 (2) ◽  
pp. 601-607 ◽  
Author(s):  
V. A. Ryzhanskii ◽  
A. G. Ivanov ◽  
A. A. Uskov ◽  
M. V. Kharlamov ◽  
V. V. Novikov ◽  
...  
Keyword(s):  
Energy Release ◽  
Reactor Core

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Tomasz Kozlowski ◽  
Hyun Chul Lee ◽  
Deokjung Lee
Keyword(s):  
Comparative Analysis ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Computer Code ◽  
Anisotropic Scattering ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Modeling Approach ◽  
Significant Difference

A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

scholarly journals Hydrodynamic planetary thermosphere model: 2. Coupling of an electron transport/energy deposition model

2008 ◽  
Vol 113 (E7) ◽  
Author(s):  
Feng Tian ◽  
Stanley C. Solomon ◽  
Liying Qian ◽  
Jiuhou Lei ◽  
Raymond G. Roble
Keyword(s):  
Electron Transport ◽  
Energy Deposition ◽  
Deposition Model

Low-Flux Neutron Radiation Detection Technology with High Sensitivity

2014 ◽  
Vol 668-669 ◽  
pp. 924-927
Author(s):  
Yang Liu ◽  
Zhen Ni Xing ◽  
Guo Zheng Zhu
Keyword(s):  
Energy Deposition ◽  
Plastic Scintillator ◽  
Detection Efficiency ◽  
Radiation Detection ◽  
High Sensitivity ◽  
Neutron Radiation ◽  
Fission Neutron ◽  
Deposition Flux ◽  
Low Intensity ◽  
Detection Technology

Boron-containing plastic scintillator detectors have a high detection efficiency for low-intensity thermal neutrons and fast neutrons which is currently the preferred types of neutron detector. This article is based on Monte Carlo method, studied boron-containing plastic scintillator for neutron detection performance, and analysis the energy deposition flux characteristics and detection efficiency when low intensity fission neutron incident to the boron plastic scintillator. We obtain the low-flux neutron detector performance in a variety of neutron source energy, boron-containing plastic scintillator diameter and length. Results showed that, when the boron-containing plastic scintillator lengths increase, the energy deposition flux will increase. When the length and diameter is constant, increasing source strength can increase the energy deposition flux brought by the recoil proton to a certain extent. When the source intensity over after thermal neutrons, due to the decrease of the cross section, the energy deposition fluxes brought by the react of neutrons and will decrease. The results provide help for low intensity fission neutron radiation detection technology with high sensitivity.

scholarly journals Study of the Reactor Control System of MSHIM in AP1000

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Xinyu Wei ◽  
Fuyu Zhao
Keyword(s):  
Control System ◽  
Power Control ◽  
Control Strategy ◽  
Power Distribution ◽  
Reactor Core ◽  
Reference Value ◽  
Engineering Practice ◽  
Mechanism Analysis ◽  
Average Temperature ◽  
Power Control System

According to the mechanism analysis and simulation of power control system of MSHIM in AP1000, a modified MSHIM (Mechanical Shim) control strategy is presented, which employs the error between the reactor coolant average temperature and its reference value as the unique control signal with a P-controller added. The modified MSHIM control strategy is verified by simulations of three typical working conditions. The results show that the modified power control system satisfies the needs of reactor core power control and power distribution control. The conclusions have reference value for the engineering practice.

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