Neutron and Neutron-Breeding Medium Interaction Process Description Within the Physical Birth-and-Death Model

2020 ◽  
Vol 23 (4) ◽  
pp. 428-434
Author(s):  
T. N. Korbut ◽  
M. V. Bobkova ◽  
E. A. Rudak ◽  
I. A. Zubets
Keyword(s):  
Physical Interpretation ◽  
Nuclear Reactor ◽  
Kinetic Characteristic ◽  
Interaction Process ◽  
Current Interest ◽  
Mathematical Apparatus ◽  
Reactor Physics ◽  
New Approach ◽  
Analytic Methods ◽  
Physics Problems

Analytic methods for nuclear reactor physics problems are still of current interest. They allow the physical interpretation to be obtained for studied processes within simple mathematical apparatus. This work proposes a new approach of neutron and neutronbreeding medium interaction process description based on birth-and-death model. This approach was named the physical birth-and-death model. The equations for the main kinetic characteristic are presented and reactivity values are estimated for two subcritical nuclear assemblies: MASURCA and KUCA.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

scholarly journals Advances in Monte-Carlo code TRIPOLI-4®’s treatment of the electromagnetic cascade

2018 ◽  
Vol 170 ◽  
pp. 01008
Author(s):  
Davide Mancusi ◽  
Alice Bonin ◽  
François-Xavier Hugot ◽  
Fadhel Malouch
Keyword(s):  
Monte Carlo ◽  
Nuclear Reactor ◽  
Thick Target ◽  
Monte Carlo Code ◽  
Reactor Physics ◽  
Validation And Verification ◽  
Electron Positron ◽  
A Charge ◽  
Charge Deposition ◽  
Nuclear Instrumentation

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

scholarly journals Virtual Nuclear Reactor for Education of Nuclear Reactor Physics

2008 ◽  
Vol 28-1 (2) ◽  
pp. 1191-1191
Author(s):  
Masashi TSUJI ◽  
Takashi NARABAYASHI ◽  
Youichiro SHIMAZU
Keyword(s):  
Nuclear Reactor ◽  
Reactor Physics

Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation

2018 ◽  
Author(s):  
Hany S. Abdel-Khalik ◽  
Dongli Huang ◽  
Ondrej Chvala ◽  
G. Ivan Maldonado
Keyword(s):  
Cross Sections ◽  
Degrees Of Freedom ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Reactor Physics ◽  
Rigorous Approach ◽  
Uncertainty Space ◽  
Core Simulation

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

scholarly journals Influence of operating firing field technologies on operational performance parameters

2018 ◽  
Vol 216 ◽  
pp. 02028
Author(s):  
Aleksandr Kazakov ◽  
Anton Suprunovskiy
Keyword(s):  
Control System ◽  
Computer Modeling ◽  
Operational Performance ◽  
Queuing Systems ◽  
Performance Parameters ◽  
Mathematical Apparatus ◽  
Railway Transport ◽  
New Approach ◽  
Software Product ◽  
Field Control

At present, introduction of operating firing field technologies is one of the urgent tasks of the Russian railway transport, as their use can significantly improve operational performance parameters. The factors and conditions of transition to operating firing field control system for operational performance are investigated in this paper. It provides the analysis of the problems that arise during the work of the operating firing field and proposes measures for their elimination. Methods of mathematical and computer modeling using the author’s software product were used. A new approach to multilevel simulation of the work of a transport operating firing field has been developed, which involves the use of the mathematical apparatus of multiphase queuing systems.

Transition Region of Nuclear Vessel Steels: Master Curve Approach Using Small Punch Notched Specimens

2017 ◽  
Vol 734 ◽  
pp. 77-86
Author(s):  
Roberto Lacalle ◽  
David Andrés ◽  
José Alberto Álvarez ◽  
Federico Gutiérrez-Solana
Keyword(s):  
Transition Region ◽  
Master Curve ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Vessel Steel ◽  
New Approach ◽  
Small Punch ◽  
Impact Tests ◽  
Integrity Assessments

The behaviour of the materials in ductile-brittle transition region must be known when performing structural integrity assessments of nuclear reactor vessels working under the effects of neutron irradiation. The characterisation of this region has been usually carried out by means of Charpy impact tests. Just during last few years new approaches based on direct fracture mechanics tests have begun to be used. In most of these cases, the Master Curve methodology, which allows the transition region to be characterised using only one parameter (T0 reference temperature), has been employed. In this paper the transition region of two materials –one vessel steel and one common structural steel-has been characterised by means of Small Punch Tests. First of all, this zone has been characterised using conventional specimens and the results were compared with those of Charpy impact tests. Finally a new approach based on the use of notched Small Punch samples together with Master Curve methodology has been proposed.

scholarly journals Elementary Introduction to Nuclear Reactor Physics

Physics Today ◽  
1962 ◽  
Vol 15 (2) ◽  
pp. 58-58
Author(s):  
S. E. Liverhant ◽  
Jacques Romain
Keyword(s):  
Nuclear Reactor ◽  
Reactor Physics
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