scholarly journals EFFECT OF ENERGY DEPOSITION MODELLING IN COUPLED STEADY STATE MONTE CARLO NEUTRONICS/THERMAL HYDRAULICS CALCULATIONS

2021 ◽  
Vol 247 ◽  
pp. 06001
Author(s):  
Riku Tuominen ◽  
Ville Valtavirta ◽  
Manuel García ◽  
Diego Ferraro ◽  
Jaakko Leppänen
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Power Distribution ◽  
Energy Deposition ◽  
Computational Time ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
New Treatment ◽  
Effective Multiplication

In coupled calculations with Monte Carlo neutronics and thermal hydraulics the Monte Carlo code is used to produce a power distribution which in practice means tallying the energy deposition. Usually the energy deposition is estimated by making a simple approximation that energy is deposited only in fission reactions. The goal of this work is to study how the accuracy of energy deposition modelling affects the results of steady state coupled calculations. For this task an internal coupling between Monte Carlo transport code Serpent 2 and subchannel code SUBCHANFLOW is used along with a recently implemented energy deposition treatment of Serpent 2. The new treatment offers four energy deposition modes each of which offers a different combination of accuracy and required computational time. As a test case, a 3D PWR fuel assembly is modelled with different energy deposition modes. The resulting effective multiplication factors are within 30 pcm. Differences of up to 100K are observed in the fuel temperatures.

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.

Newly Developed Coupling Scheme in RMC With Internal Thermal Feedback and OTF Doppler-Broadening Method

2017 ◽  
Author(s):  
Ouwen Yexin ◽  
Shanfang Huang ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Cross Section ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Simulation Analysis ◽  
Coupling Scheme ◽  
Monte Carlo Code ◽  
Doppler Broadening ◽  
Thermal Hydraulics

RMC (Reactor Monte Carlo)[1] is a self-developed Monte Carlo code for nuclear reactor analysis by Reactor Engineering Analysis Lab (REAL), Tsinghua University. On the basis of the self-developed subchannel module (RMC-TH) and Monte Carlo Cell Tally, the internal coupling interface is developed, which combines both input files to one and realizes the fast mesh correspondence process using the cell expansion technology for repeated structure with thermal-hydraulics feedback. It breaks through the bottleneck of geometrical extensibility for coupled code. On-the-fly Doppler broadening method is adopted as the way to consider the temperature effect on microscopic cross section, which only needs the 0 K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on PWR fuel pin and 17×17 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology. Therefore, a promising technology roadmap for the large-scale and geometrically universal nuclear reactor in both steady-state and transient conditions with thermal-hydraulic feedback are established. The roadmap can be further applied to neutronics-thermal-hydraulics-depletion coupling in multi-physics simulation process.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Kerntechnik ◽  
2021 ◽  
Vol 86 (4) ◽  
pp. 302-311
Author(s):  
M. E. Korkmaz ◽  
N. K. Arslan
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Study ◽  
Minor Actinides ◽  
Monte Carlo Code ◽  
Effective Multiplication Factor ◽  
Conceptual Core ◽  
Effective Multiplication

Abstract Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

The Modelling and Coupling Methodology of ANSYS CFX Using Porous Media for PB-AHTR

2013 ◽  
Author(s):  
Linsen Li ◽  
Haomin Yuan ◽  
Kan Wang
Keyword(s):  
Porous Media ◽  
Monte Carlo ◽  
Steady State ◽  
High Temperature ◽  
Monte Carlo Code ◽  
First Principle ◽  
Pebble Bed ◽  
Coupled Calculation ◽  
Ansys Cfx ◽  
High Temperature Reactor

This paper introduces a first-principle steady-state coupling methodology using the Monte Carlo Code RMC and the CFD code CFX which can be used for the analysis of small and medium reactors. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. A Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core is modeled using this method. The porous media is used in the CFX model to simulate the pebble bed structure in PB-AHTR. This research concludes that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations.

Validating the Serpent-Ants Calculation Chain Using BEAVRS Fresh Core HZP Data

2021 ◽  
Author(s):  
Ville Valtavirta ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Monte Carlo ◽  
Best Practices ◽  
Power Distribution ◽  
Measured Data ◽  
Monte Carlo Code ◽  
Generation Process ◽  
Isothermal Temperature ◽  
Control Rod ◽  
Temperature Coefficients ◽  
Group Constant

Abstract The Serpent Monte Carlo code and the Serpent-Ants two step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod group worths and isothermal temperature coefficients are compared between Serpent and Serpent-Ants as well as against the experimental measurements. Furthermore, radial power distributions in the unrodded and rodded core configurations are compared between Serpent and Serpent-Ants. In addition to providing results using a best practices calculation chain, the effects of several simplifications or omissions in the group constant generation process on the results are estimated. Both the direct and two-step neutronics solutions provide results close to the measured values. Comparison between the measured data and the direct Serpent Monte Carlo solution yields RMS differences of 12.1 mg/kg, 25.1 × 10-5 and 0.67 × 10-5 K-1 for boron, control rod worths and temperature coefficients respectively. The two-step Serpent-Ants solution reaches a similar level of accuracy with RMS differences of 17.4 mg/kg, 23.6 × 10-5 and 0.29 × 10-5 K-1. The match in the radial power distribution between Serpent and Serpent-Ants was very good with the RMS and maximum for pin power errors being 1.31 % and 4.99 % respectively in the unrodded core and 1.67 %(RMS) and 8.39 % (MAX) in the rodded core.

Coupled neutronics–fuel behavior calculations in steady state using the Serpent 2 Monte Carlo code

2017 ◽  
Vol 100 ◽  
pp. 50-64 ◽  
Author(s):  
Ville Valtavirta ◽  
Jaakko Leppänen ◽  
Tuomas Viitanen
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Monte Carlo Code ◽  
Fuel Behavior

scholarly journals Beryllium erosion and redeposition in ITER H, He and D-T discharges

2021 ◽  
Author(s):  
Juri Romazanov ◽  
Andreas Kirschner ◽  
Sebastijan Brezinsek ◽  
Richard A Pitts ◽  
Dmitriy V. Borodin ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Steady State ◽  
Flow Velocity ◽  
Helium Plasma ◽  
Dust Formation ◽  
Monte Carlo Code ◽  
Main Chamber ◽  
First Wall ◽  
Erosion Rates ◽  
Hydrogen Deuterium

Abstract The Monte-Carlo code ERO2.0 was used to simulate steady-state erosion and transport of beryllium (Be) in the ITER main chamber. Various plasma scenarios were tested, including a variation of the main species (hydrogen, deuterium, helium), plasma conditions (density, temperature, flow velocity) and magnetic configurations. The study provides valuable predictions for the Be transport to the divertor, where it is expected to be an important contributor to dust formation and fuel retention due to build-up of co-deposited layers. The Be gross and net erosion rates provided by this study can help identifying first wall regions with potentially critical armour lifetime.

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.

Sign in / Sign up

Export Citation Format

Share Document