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

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.

scholarly journals MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4®

2021 ◽  
Vol 247 ◽  
pp. 07019 ◽  
Author(s):  
Margaux Faucher ◽  
Davide Mancusi ◽  
Andrea Zoia
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Coolant Temperature ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
Transient Simulations ◽  
Neutron Power

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed.

The Mean Projected Range and Range Straggling of Nd Ions Implanted in Silicon Carbide

2011 ◽  
Vol 474-476 ◽  
pp. 565-569
Author(s):  
Xi Feng Qin ◽  
Shuang Li ◽  
Feng Xiang Wang ◽  
Yi Liang
Keyword(s):  
Silicon Carbide ◽  
Monte Carlo ◽  
Spectrum Analysis ◽  
Monte Carlo Code ◽  
Experiment Data ◽  
Lateral Deviation ◽  
Projected Range ◽  
The Mean ◽  
The Difference ◽  
Good Agreement

In view of the influence of the projected range, the range straggling, and the lateral deviation of ions in materials on the property of device in the fabrication of photoelectric integration devices by ion implantation, the mean projected ranges and range straggling for energetic 200 – 500 keV Nd ions implanted in 6H-SiC were measured by means of Rutherford backscattering followed by spectrum analysis. The measured values are compared with Monte Carlo code (SRIM2006) calculations. It has been found that the measured values of the mean projected range Rp are good agreement with the SRIM calculated values; for the range straggling △Rp, the difference between the experiment data and the calculated results is much higher than that of Rp

scholarly journals Portable LQCD Monte Carlo code using OpenACC

2018 ◽  
Vol 175 ◽  
pp. 09008
Author(s):  
Claudio Bonati ◽  
Enrico Calore ◽  
Simone Coscetti ◽  
Massimo D’Elia ◽  
Michele Mesiti ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Parallel Programming ◽  
Scientific Computing ◽  
State Of The Art ◽  
Production Level ◽  
Monte Carlo Code ◽  
Design And Optimization ◽  
Code Changes ◽  
Many Core ◽  
Descriptive Level

Varying from multi-core CPU processors to many-core GPUs, the present scenario of HPC architectures is extremely heterogeneous. In this context, code portability is increasingly important for easy maintainability of applications; this is relevant in scientific computing where code changes are numerous and frequent. In this talk we present the design and optimization of a state-of-the-art production level LQCD Monte Carlo application, using the OpenACC directives model. OpenACC aims to abstract parallel programming to a descriptive level, where programmers do not need to specify the mapping of the code on the target machine. We describe the OpenACC implementation and show that the same code is able to target different architectures, including state-of-the-art CPUs and GPUs.

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 Comparative Analysis of Reactivity Calculations for the CERMET Fueled ADS with Serpent and MCNP6 Codes

2021 ◽  
Vol 37 (1) ◽  
Author(s):  
Vu Thanh Mai ◽  
Donny Hartanto ◽  
Pham Nhu Viet Ha ◽  
Nguyen Thi Dung ◽  
Bui Thi Hoa ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Comparative Analysis ◽  
Nuclear Waste ◽  
Inherent Safety ◽  
Fuel Temperature ◽  
Safety Feature ◽  
Temperature Feedback ◽  
Driven System ◽  
Accelerator Driven System ◽  
Good Agreement

The ADS (accelerator driven system) is recognized as a promising system to annihilate the radioactivity of nuclear waste with its inherent safety feature and waste transmutation potential. Thus, conceptual designs of ADS are widely carrying out. In order to verify the accuracy of an innovative ADS core modeling by using simulation codes, the reactivity calculations of CERMET fueled ADS were conducted using two Monte Carlo codes, Serpent and MCNP6 with ENDF/B-VII.0 library. The comparison shows a good agreement between two codes including the eigenvalue (less than 50 pcm) and fuel temperature feedback (discrepancy is within the standard deviation). It implies that the ADS was modelled successfully and can be used for further investigation.  Keywords: CERMET fueled ADS, Serpent code, MCNP6, reactivity calculation.

scholarly journals Switchable DBR Filters Using Semiconductor Distributed Doped Areas (ScDDAs)

Electronics ◽  
2020 ◽  
Vol 9 (12) ◽  
pp. 2021
Author(s):  
Rozenn Allanic ◽  
Denis Le Berre ◽  
Cédric Quendo ◽  
David Chouteau ◽  
Virginie Grimal ◽  
...  
Keyword(s):  
High Frequency ◽  
Degrees Of Freedom ◽  
State Of The Art ◽  
Active Surface ◽  
The State ◽  
Semiconductor Substrate ◽  
Great Flexibility ◽  
Active Elements ◽  
Insertion Losses ◽  
Good Agreement

This paper presents a novel way to switch dual-behavior resonator (DBR) filters without any additional active surface-mount components. By using a semiconductor substrate, we were able to simultaneously co-design the filters and semiconductor distributed doped areas (ScDDAs) with integrated N+PP+ junctions as active elements. These ScDDAs act as electrical vias in the substrate, which makes it possible to have an open-circuited resonator in the OFF state and a short-circuited resonator in the ON state, and, consequently, to control the transmission zeroes of the filters. This method offers degrees of freedom as the dimensions and positions of these doped areas can be chosen to obtain the best performances. In this study, four filters were simulated and fabricated to spotlight different possibilities for the dimensions and positions of the ScDDA to control the low- or high-frequency transmission zero of the filters. The simulations were in very good agreement with the measured results. All the filters present insertion losses lower than 2 dB in the OFF and ON states, a great flexibility in the frequency choice, and good agility compared with the state of the art.

scholarly journals Processing and application of nuclear data for low temperature criticality assessment

2020 ◽  
Vol 239 ◽  
pp. 14006
Author(s):  
Tim Ware ◽  
David Hanlon ◽  
Tara Hanlon ◽  
Richard Hiles ◽  
Malcolm Lingard ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Low Temperature ◽  
Cross Sections ◽  
Nuclear Data ◽  
Scattering Data ◽  
Monte Carlo Code ◽  
Temperature Range ◽  
Nuclear Data Library ◽  
Criticality Assessment ◽  
Good Agreement

Until recently, criticality safety assessment codes had a minimum temperature at which calculations can be performed. Where criticality assessment has been required for lower temperatures, indirect methods, including reasoned argument or extrapolation, have been required to assess reactivity changes associated with these temperatures. The ANSWERS Software Service MONK® version 10B Monte Carlo criticality code, is capable of performing criticality calculations at any temperature, within the temperature limits of the underlying nuclear data in the BINGO continuous energy library. The temperature range of the nuclear data has been extended below the traditional lower limit of 293.6 K to 193 K in a prototype BINGO library, primarily based on JEFF-3.1.2 data. The temperature range of the thermal bound scattering data of the key moderator materials was extended by reprocessing the NJOY LEAPR inputs used to produce bound data for JEFF-3.1.2 and ENDF/B-VIII.0. To give confidence in the low temperature nuclear data, a series of MONK and MCBEND calculations have been performed and results compared against external data sources. MCBEND is a Monte Carlo code for shielding and dosimetry and shares commonalities to its sister code MONK including the BINGO nuclear data library. Good agreement has been achieved between calculated and experimental cross sections for ice, k-effective results for low temperature criticality benchmarks and calculated and experimentally determined eigenvalues for thermal neutron diffusion in ice. To quantify the differences between ice and water bound scattering data a number of MONK criticality calculations were performed for nuclear fuel transport flask configurations. The results obtained demonstrate good agreement with extrapolation methods. There is a discernible difference in the use of ice and water data.

Comparison of the Reactivity Effects Calculated by DRAGON and Serpent for a PHWR 37-Element Fuel Bundle

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Jawad Haroon ◽  
Leslie Kicka ◽  
Subhramanyu Mohapatra ◽  
Eleodor Nichita ◽  
Peter Schwanke
Keyword(s):  
Monte Carlo ◽  
Heavy Water ◽  
Monte Carlo Methods ◽  
Statistical Error ◽  
Natural Uranium ◽  
Monte Carlo Code ◽  
Fuel Temperature ◽  
Fuel Bundle ◽  
Deterministic Methods ◽  
Heavy Water Reactor

Deterministic and Monte Carlo methods are regularly employed to conduct lattice calculations. Monte Carlo methods can effectively model a large range of complex geometries and, compared to deterministic methods, they have the major advantage of reducing systematic errors and are computationally effective when integral quantities such as effective multiplication factor or reactivity are calculated. In contrast, deterministic methods do introduce discretization approximations but usually require shorter computation times than Monte Carlo methods when detailed flux and reaction-rate solutions are sought. This work compares the results of the deterministic code DRAGON to the Monte Carlo code Serpent in the calculation of the reactivity effects for a pressurized heavy water reactor (PHWR) lattice cell containing a 37-element, natural uranium fuel bundle with heavy water coolant and moderator. The reactivity effects are determined for changes to the coolant, moderator, and fuel temperatures and to the coolant and moderator densities for zero-burnup, mid-burnup [3750  MWd/t(U)] and discharge burnup [7500  MWd/t(U)] fuel. It is found that the overall trend in the reactivity effects calculated using DRAGON match those calculated using Serpent for the burnup cases considered. However, differences that exceed the amount attributable to statistical error have been found for some reactivity effects, particularly for perturbations to coolant and moderator density and fuel temperature.

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