| 1. |
- Demaziere, Christophe, 1973, et al.
(author)
-
Development and test of a new verification scheme for transient core simulators
- 2017
-
In: Transactions of the American Nuclear Society. - 0003-018X. ; 116, s. 1025-1026
-
Conference paper (peer-reviewed)abstract
- Transient calculations in commercial nuclear reactors are performed while typically relying on a time-dependent neutron transport solver or a low-order solver (i.e. diffusion). In order to be licensed, the codes used by the industry need to go through a process of verification and validation, with the verification carried out by comparing the results of simulations to analytical or semi-analytical solutions. Such analytical or semi-analytical solutions can only be obtained if the system to be modelled during the verification process is either fully homogeneous or piece-wise homogeneous.This paper reports on the development of a different verification approach that can be applied to fully heterogeneous systems. It relies on the extraction of the point-kinetic response of the reactor (which can be estimated from the results of core simulations) and on its subsequent comparison with its expected analytical form.
|
|
| 2. |
- Demaziere, Christophe, 1973, et al.
(author)
-
Development of a point-kinetic verification scheme for nuclear reactor applications
- 2017
-
In: Journal of Computational Physics. - : Elsevier BV. - 1090-2716 .- 0021-9991. ; 339, s. 396-411
-
Journal article (peer-reviewed)abstract
- In this paper, a new method that can be used for checking the proper implementation of time- or frequency-dependent neutron transport models and for verifying their ability to recover some basic reactor physics properties is proposed. This method makes use of the application of a stationary perturbation to the system at a given frequency and extraction of the point-kinetic component of the system response. Even for strongly heterogeneous systems for which an analytical solution does not exist, the point-kinetic component follows, as a function of frequency, a simple analytical form. The comparison between the extracted point-kinetic component and its expected analytical form provides an opportunity to verify and validate neutron transport solvers. The proposed method is tested on two diffusion-based codes, one working in the time domain and the other working in the frequency domain. As long as the applied perturbation has a non-zero reactivity effect, it is demonstrated that the method can be successfully applied to verify and validate time- or frequency-dependent neutron transport solvers. Although the method is demonstrated in the present paper in a diffusion theory framework, higher order neutron transport methods could be verified based on the same principles.
|
|
| 3. |
- Demazière, C., et al.
(author)
-
Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
- 2015
-
In: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
-
Journal article (peer-reviewed)abstract
- This paper presents the development of a numerical tool meant at modelling the effect of stationary fluctuations in nuclear cores for systems cooled with either liquid water or boiling water. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool then determines in the frequency domain the three-dimensional distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the Homogeneous Equilibrium Model, with a void fraction correction based on a pre-computed distribution of the static slip ratio (when two-phase flow conditions are encountered). Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool, currently entirely Matlab based, requires minimal input data, mostly in form of the three-dimensional distributions of the macroscopic cross-sections and their relative dependence on coolant density and fuel temperature, the point-kinetic parameters of the core, as well as the three-dimensional distribution of the slip ratio (in case of two-phase flow conditions) and of the heat transfer coefficient. Such data can be provided by any static core simulator that thus needs to be run prior to using the present tool. In addition to briefly summarizing the different test cases used to verify the code, the paper also presents the results of simulations performed for a typical Pressurized Water Reactor and for a typical Boiling Water Reactor, as illustrations of the capabilities of the tool.
|
|
| 4. |
- Demaziere, Christophe, 1973, et al.
(author)
-
Estimation of the zero-power reactor transfer from a 3-dimensional core simulator in the frequency domain
- 2016
-
In: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016.
-
Conference paper (peer-reviewed)abstract
- It is well known in reactor dynamics that the so-called open-loop or zero-power reactor transfer function, which assumes a point-kinetic behavior of the system, has a simple analytical expression in the frequency domain. This expression depends on the effective fraction of delayed neutrons, the decay constant of the precursors of delayed neutrons, and the neutron mean generation time. In this paper, a methodology is proposed to recover the point-kinetic component of the fluctuations in neutron flux induced by perturbations of macroscopic cross-sections. These fluctuations can be estimated by any open-loop reactor simulator working in the frequency domain, and the proposed method could thus be used as a means to validate the simulator against the theoretical expression of the transfer function. This validation exercise represents one of the very few cases where the response of a heterogeneous core can be compared to the evaluation of an analytical expression. In this paper, the methodology is also demonstrated using the CORE SIM tool in two test situations: a localized absorber of variable strength, and a travelling perturbation. In both cases, the simulator is able to reproduce the expected frequency-dependence of the reactor transfer function, despite the fact that the reactor response significantly deviates from point-kinetic for localized perturbations at high frequencies. It has nevertheless to be pointed out that the proposed method only works if the applied perturbation has a non-zero reactivity effect.
|
|
| 5. |
- Demaziere, Christophe, 1973, et al.
(author)
-
Modelling of stationary fluctuations in nuclear reactor cores in the frequency domain
- 2015
-
In: Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015. - : American Nuclear Society. - 9781510808041 ; , s. 2406-2419
-
Conference paper (peer-reviewed)abstract
- This paper presents the development of a numerical tool to simulate the effect of stationary fluctuations in Light Water Reactor cores. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool calculates the three-dimensional space-frequency distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the homogeneous equilibrium model complemented with pre-computed static slip ratio. Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool is currently entirely Matlab based with input data provided by an external static core simulator. The paper also presents the results of dynamic simulations performed for a typical pressurized water reactor and for a typical boiling water reactor, as illustrations of the capabilities of the tool.
|
|
| 6. |
- Dykin, Victor, 1985, et al.
(author)
-
Predictive BWR core stability using feedback reactivity coefficients projected on neutronic eigenmodes
- 2019
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 124, s. 1-8
-
Journal article (peer-reviewed)abstract
- The determination of the stability properties of Boiling Water Reactors usually rely on performing many time-dependent calculations for various combinations of values for the core power and the core flow. The aim of such calculations is to estimate the variation of the Decay Ratio in the core power/flow operating map, from which possible exclusion areas are defined. This paper demonstrates using a Reduced Order Model that the stability properties of a core with respect to global and regional oscillations are entirely determined by the projection of the feedback reactivity coefficients onto pairs of neutronic eigenmodes and their adjoint functions. This means that such projections inherently contain all information about the stability properties and their examination is sufficient to characterize the stability of a core. Most notably, the relative contributions of each fuel assembly to the core-wise projections give an indication to the core designer about the fuel assemblies possibly destabilizing the core. The core designer could thereafter improve core stability by either moving such assemblies to other locations or use another fuel assembly design. Although the method could be used independently of detailed stability calculations, the approach detailed in this study provides a more qualitative than quantitative core stability evaluation. This means that the method is most efficient if the stability features of a reference core are known.
|
|
| 7. |
- Dykin, Victor, 1985, et al.
(author)
-
Remark on the neutron noise induced by propagating perturbations in an MSR
- 2016
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 90, s. 93-105
-
Journal article (peer-reviewed)abstract
- The neutron noise induced by propagating perturbations in a simple model of a Molten Salt Reactor (MSR) is calculated and analyzed using one/two-group diffusion theory. The novelty, as compared to previous works, is that the noise source includes also the fluctuations of the fission cross sections and the fluid velocity, in addition to the previous case when only the fluctuations of the absorption cross section were accounted for. Another novelty is that the solution is obtained through the matrix Green's function of the flux and precursor equations, these two being kept separate. Inclusion of each of these two new noise sources leads to a structure of the noise source, and hence also that of the neutron noise, which is conceptually different from the case when only the fluctuations of the absorption cross sections are treated, with some surprising features. The use of the matrix Green's function is advantageous to understand the new features, and it helps to point out some new aspects of the neutron noise even in traditional systems, which have not been noticed before. The results contribute to the understanding and interpretation of the neutron noise in MSRs. (C) 2015 Elsevier Ltd. All rights reserved.
|
|
| 8. |
- Dykin, Victor, 1985, et al.
(author)
-
Ringhals Diagnostics and Monitoring, Annual Research Report 2015
- 2015
-
Reports (other academic/artistic)abstract
- This report gives an account of the work performed by the Division of Subatomic and Plasma Physics (former Division of Nuclear Engineering), Chalmers, in the frame of a research collaboration with Ringhals, Vattenfall AB, contract No. 630217-031. The contract constitutes a 1-year co-operative research work concerning diagnostics and monitoring of the BWR and PWR units. The work in the contract has been performed between January 1st 2015, and December 31st, 2015. During this period, we have worked with five main items as follows: 1. Development of the mode separation model with an extension to describe 3-D core barrel vibrations; 2. Analysis of new ex-core measurements, taken in R-4 after power uprate; 3. Investigation of the correctness of the hypothesis that the reactivity component extracted from the ex-core detector signals can be due to fuel assembly vibrations with CORE SIM; 4. A basic study in neutron noise theory which could provide some indirect support for the determination of the void fraction from neutron noise measurements; 5. A preliminary study of the possibility of modelling 3-dimensional fuel assembly vibrations in a realistic PWR system with the CORE SIM simulator. This work was performed at the Nuclear Engineering Group of the Division of Subatomic and Plasma Physics, Chalmers University of Technology by Victor Dykin (project co-ordinator), Cristina Montalvo (visitor from the Technical University of Madrid), Hoai-Nam Tran (research collaborator from Duy Tan University), Imre Pázsit and Henrik Nylén, who was also the contact person at Ringhals.
|
|
| 9. |
|
|
| 10. |
- Dykin, Victor, 1985, et al.
(author)
-
The Molten Salt Reactor Point-Kinetic Component of Neutron Noise in Two-Group Diffusion Theory
- 2016
-
In: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 193:3, s. 404-415
-
Journal article (peer-reviewed)abstract
- The derivation of the point-kinetic component of the neutron noise in two-group diffusion theory in molten salt reactors (MSRs), based on different techniques, is discussed. First, the point-kinetic component is calculated by projecting the corresponding full space-frequency-dependent solution onto the static adjoint. Then, following the standard procedure in reactor physics, the point-kinetic solution is determined by solving the linearized point-kinetic equations. Both results are thereafter analyzed and compared quantitatively. Such a comparison clearly indicates that the solution obtained by the conventional derivation, i.e., from the point-kinetic equations, significantly differs from the exact one and is not able to reproduce certain features of the latter. Similar discrepancies between the two methods were also pointed out and confirmed earlier in one-group MSR calculations.
|
|
