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.

Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

2015 ◽  
Vol 281 ◽  
pp. 58-71 ◽  
Author(s):  
M.L. Jayalal ◽  
Suja Ramachandran ◽  
S. Rathakrishnan ◽  
S.A.V. Satya Murty ◽  
M. Sai Baba
Keyword(s):  
Genetic Algorithm ◽  
Heavy Water ◽  
Water Reactor ◽  
Fuel Bundle ◽  
Heavy Water Reactor

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 Fully quantum calculation of the second and third virial coefficients of water and its isotopologues from ab initio potentials

2018 ◽  
Vol 212 ◽  
pp. 467-497 ◽  
Author(s):  
Giovanni Garberoglio ◽  
Piotr Jankowski ◽  
Krzysztof Szalewicz ◽  
Allan H. Harvey
Keyword(s):  
Monte Carlo ◽  
Heavy Water ◽  
Path Integral ◽  
Monte Carlo Methods ◽  
Virial Coefficients ◽  
Quantum Calculation ◽  
Path Integral Monte Carlo ◽  
The Third ◽  
Ab Initio Potentials ◽  
State Of Art

Path-Integral Monte Carlo methods were applied to calculate the second, B(T), and the third, C(T), virial coefficients for water and heavy water from state-of-art flexible potentials.

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.

scholarly journals ON THE USE OF A CENTRAL THORIUM FUEL ELEMENT IN PRESSURE-TUBE HEAVY-WATER REACTOR FUEL BUNDLES

2018 ◽  
Vol 7 (2) ◽  
pp. 147-155 ◽  
Author(s):  
Michael McDonald ◽  
Megan Moore ◽  
Dan Wojtaszek ◽  
Nicholas Chornoboy ◽  
Geoffrey Edwards
Keyword(s):  
Fuel Element ◽  
Heavy Water ◽  
Fuel Cycle ◽  
Natural Uranium ◽  
Pressure Tube ◽  
Slight Reduction ◽  
Energy Potential ◽  
Thorium Dioxide ◽  
Water Reactor ◽  
Heavy Water Reactor

An incremental approach to introducing thorium to the conventional pressure-tube heavy-water reactor natural uranium fuel cycle is investigated. The approach involves the replacement of the centre fuel element of the bundle with an element of thorium dioxide. Increasing the operating margin of a key safety parameter, the coolant void reactivity, is a prime motivating factor. The analyses showed that the simple use of a single pin of thorium is unlikely to be economically advantageous due to a large burnup penalty and increased fuel costs. However, a slight reduction in the void reactivity is observed, and this approach does allow the exploitation of the energy potential available in thorium as an alternative nuclear fuel resource through the development of a U-233 resource. This bundle concept may also be advantageous from a fuel disposal point of view, as the fuel requires less time in storage before emplacement in a deep geological repository.

One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation

2018 ◽  
Author(s):  
Xiaotong Shang ◽  
Guanlin Shi ◽  
Kan Wang
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Cross Sections ◽  
Monte Carlo Code ◽  
Event Analysis ◽  
Step Method ◽  
Kinetics Parameters ◽  
Transient Event ◽  
Continuous Energy ◽  
Deterministic Methods

The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability (IFP) method is introduced to calculate the neutron importance which is able to represent the adjoint neutron flux in continuous energy problem and have been applied to the calculation of kinetics parameters. However, the adjoint neutron flux can’t be obtained directly and applied to both Monte Carlo transient event analysis and deterministic methods. In this study, a method based on IFP is studied and implemented in Monte Carlo code RMC. The multi-group adjont neutron flux can be obtained directly through the discretization of energy and space with the modification of fission neutrons through continuous energy Monte Carlo calculations. The obtained multi-group adjoint neutron flux can be used in both Monte Carlo transient analysis and deterministic methods.

scholarly journals Study of natural uranium fuel for a new reactor design TEPLATOR

2021 ◽  
Vol 253 ◽  
pp. 07012
Author(s):  
Tomas Peltan ◽  
Eva Vilimova ◽  
Radek Skoda
Keyword(s):  
Heavy Water ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
District Heating ◽  
Alternative Fuel ◽  
Natural Uranium ◽  
Thermal Reactor ◽  
Monte Carlo Code ◽  
Fuel Assemblies ◽  
Irradiated Fuel

The TEPLATOR is a new type of nuclear reactor which the main purpose is producing heat for district heating. It is designed as a special thermal reactor with 55 fuel channels for fuel assemblies, which is moderated and cooled by heavy water and operated around atmospheric pressure. The TEPLATOR DEMO is designed for the use of irradiated fuel from PWR or BWR reactors. Using heavy water as the moderator and coolant in this reactor concept allows to use natural uranium as an alternative fuel in case that the irradiated fuel is not available for some reason. This solution is suitable because of the price of natural uranium and the absence of costly fuel enrichment. This article is focused on deeper analyses of alternative suitable fuel for TEPLATOR based on natural uranium and new fuel geometries. This work builds on previous research on alternative fuel material and geometry for the TEPLATOR. It is mainly concerned with the neutronic development of fuel assemblies, the possibility of manufacturing of developed fuel types, and optimization of fuel management and uranium consumption. This article contains predetermined candidates for suitable fuel geometries and new untested fuel geometry types with some new advantages. Finally, optimization of the whole reactor core and number of fuel channels was made in terms of increased safety and higher fuel burn-up. Presented calculations were performed by Monte Carlo code Seprent.

Sign in / Sign up

Export Citation Format

Share Document