| 1. |
- Lau, Cheuk Wah, 1985, et al.
(författare)
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Conceptual study of axial offset fluctuations upon stepwise power changes in a thorium-plutonium core to improve load-following conditions
- 2014
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 84-89
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Tidskriftsartikel (refereegranskat)abstract
- The increased share of renewable energy, such as wind and solar power, will increase the demand for load-following power sources, and nuclear reactors could be one option. However, during rapid load-following events, traditional UOX cores could be restricted by the volatile oscillation of the power distribution. Therefore, a conceptual study on stability properties of Th-MOX PWR concerning axial offset power excursion during load-following events are investigated and discussed. The study is performed in SIMULATE-3 for a realistic PWR core (Ringhals-3) at the end of cycle, where the largest amplitude of the axial offset oscillations is expected. It is shown that the Th-MOX core possesses much better stability characteristics and shorter reactor dead time compared with a traditional UOX core, and the main reasons are the lower sensitivity to perturbations in the neutron spectrum, lower xenon poisoning and lower thermal neutron flux.
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| 2. |
- Avdic, Senada, et al.
(författare)
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Item identification with a space-dependent model of neutron multiplicities and artificial neural networks
- 2023
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Ingår i: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - 0168-9002. ; 1057
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Tidskriftsartikel (refereegranskat)abstract
- A method of calculating the neutron multiplicity rates (singles, doubles and triples rates), based on transport theory, was developed by us recently. The model treats the full 3-D spatial transport and multiplication of neutrons, accounting also for the shape of the item and the spatial distribution of the source, in one-speed theory. For a given item and its source distribution, the model can predict the multiplicity rates more precisely than the point model, on which traditional neutron multiplicity counting is based. However, so far it has not been investigated how the enhanced accuracy of the calculated multiplicity rates (i.e. the solution of the direct task) can be used to estimate the parameters of interest of the measurement item, primarily the fission rate (the solution of the inverse task). Unlike for the point model, the multiplicity rates under the extended scheme can only be given numerically, as solutions of integral transport equations, and hence an analytical inversion of the formulae is not possible. In this work it is investigated how machine learning methods, primarily the use of artificial neural networks, which only need numerical values of the solution of the direct task (the multiplicity rates), can be used for this purpose. It is shown that for numerical test items containing a mixture of 239Pu and 240Pu, the fraction of the latter varying between 4% and 25%, one can extract the masses of both isotopes from a properly trained network.
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| 3. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Development and test of a new verification scheme for transient core simulators
- 2017
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Ingår i: Transactions of the American Nuclear Society. - 0003-018X. ; 116, s. 1025-1026
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Konferensbidrag (refereegranskat)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.
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| 4. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Development of a point-kinetic verification scheme for nuclear reactor applications
- 2017
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Ingår i: Journal of Computational Physics. - : Elsevier BV. - 1090-2716 .- 0021-9991. ; 339, s. 396-411
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Tidskriftsartikel (refereegranskat)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.
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| 5. |
- Demazière, C., et al.
(författare)
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Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
- 2015
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Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
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Tidskriftsartikel (refereegranskat)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.
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| 6. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Estimation of the zero-power reactor transfer from a 3-dimensional core simulator in the frequency domain
- 2016
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Ingår i: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016.
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Konferensbidrag (refereegranskat)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.
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| 7. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Modelling of stationary fluctuations in nuclear reactor cores in the frequency domain
- 2015
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Ingår i: 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
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Konferensbidrag (refereegranskat)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.
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| 8. |
- Dykin, Victor, 1985, et al.
(författare)
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Description of the models and algorithms used in the coupled CORE SIM neutronic and thermo-hydraulic tool
- 2014
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The development of an innovative coupled neutronic/thermo-hydraulic tool is reported hereafter. The novelty of the tool resides in its versatility, since many different systems can be investigated and different kinds of calculations can be performed. More precisely, both critical systems and subcritical systems with an external neutron source can be studied, static and dynamic cases in the frequency domain (i.e. for stationary fluctuations) can be considered. For each situation, the three dimensional distributions of static neutron fluxes, all thermo-hydraulic parameters, their respective first-order noise are estimated, as well as the effective multiplication factor of the system. The main advantages of the tool, which is entirelyMATLAB based, lie with the robustness of the implemented numerical algorithms, its high portability between different computer platforms and operative systems, and finally its ease of use since no input deck writing is required. The present version of the tool, which is based on two-group diffusion theory, is mostly suited to investigate thermal systems, both Pressurized and BoilingWater Reactors (PWR and BWR, respectively). This report describes the neutronic and thermo-hydraulic models, their coupling and numerical algorithms implemented in the tool, whereas the demonstration of the tool is reported in a companion report. The tool, for which a complete user’smanual exists, is freely available on direct request to the authors of the present report.
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| 9. |
- Dykin, Victor, 1985, et al.
(författare)
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Development of a fully-consistent Reduced Order Model to study instabilities in Boiling Water Reactors
- 2012
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Ingår i: Proc. Int. Conf. on Advances in Reactor Physics – Linking Research, Industry, and Education (PHYSOR 2012), Knoxville, TN, USA, April 15-20, 2012, American Nuclear Society. - 9781622763894 ; 1, s. 332 - 345
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Konferensbidrag (refereegranskat)abstract
- A simple nonlinear Reduced Order Model to study global, regional and local instabilities in Boiling Water Reactors is described. The ROM consists of three submodels: neutron-kinetic, thermal-hydraulic and heat-transfer models. The neutron-kinetic model allows representing the time evolution of the three first neutron kinetic modes: the fundamental, the first and the second azimuthal modes. The thermal-hydraulic model describes four heated channels in order to correctly simulate out-of-phase behavior. The coupling between the different submodels is performed via both void and Doppler feedback mechanisms. After proper spatial homogenization, the governing equations are discretized in the time-domain. Several modifications, compared to other existing ROMs, have been implemented, and are reported in this paper. One novelty of the ROM is the inclusion of both azimuthal modes, which allows to study combined instabilities (in-phase and out-of-phase), as well as to investigate the corresponding interference effects between them. The second modification concerns the precise estimation of so-called reactivity coefficients or C*^{V,D}_{mn} - coefficients by using direct cross-section data from SIMULATE-3 combined with the CORE SIM core simulator in order to calculate eigenmodes. Furthermore, a non-uniform two-step axial power profile is introduced to simulate the separate heat production in the single and two-phase regions, respectively. An iterative procedure was developed to calculate the solution to the coupled neutron-kinetic/thermal-hydraulic static problem prior to solving the time-dependent problem. Besides, the possibility of taking into account the effect of local instabilities is demonstrated in a simplified manner. The present ROM is applied to the investigation of an actual instability that occurred at the Swedish Forsmark-1 BWR in 1996/1997. The results generated by the ROM are compared with real power plant measurements performed during stability tests and show a good qualitative agreement. The present study provides some insight in a deeper understanding of the physical principles which drive both core-wide and local instabilities.
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| 10. |
- Dykin, Victor, 1985, et al.
(författare)
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Development of a Reduced Order Model and its application to the Forsmark-1 Instability Event of 1996/1997
- 2010
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report describes the development of a reduced order model (ROM) which is then used to analyze a specific instability event . The ROM consists of three sub-models: a neutron-kinetic (NK) model (describing neutron transport), a thermal hydraulic (TH) model (HT) (describing the coolant flow ) and a heat transfer model (describing heat transfer between the fuel and the coolant). All these three models are coupled to each other, using two feedback mechanisms: void fraction feedback and doppler feedback. Each of the sub-models is described by a set of reduced ordinary differential equations, derived from the corresponding time-space dependent partial differential equations by using different types of approximations and mathematical methods. The neutron kinetic model is derived from the two-group time-space-dependent diffusion equations with one effective group of delayed neutrons by using eigenmode expansion. In the NK model, only the effect of the first three modes, namely the fundamental the first and the second azimuthal modes are taken into account. The thermal hydraulic model is derived from the space-time dependent mass, momentum and enthalpy local conservation equations, using spatial quadratic approximation for both the enthalpy and the quality distributions, after applying the weighted residual procedure. The equations are written for the single and for the two phase regions, separately. For the sake of simplicity, in the latter case the HEM is used. The heat transfer model is derived from an energy balance equation written for one fuel rod with three radial regions. The reduction of the (HT) equations is performed by assuming a piece-wise quadratic approximation for the fuel pellet temperature and using a weighted residual procedure (WRP). In order to have proper representation of both azimuthal modes, a four heated channels model was constructed. The recirculation loop model was also introduced into the ROM. The coupling reactivity coefficients for both void fraction and fuel temperature were calculated explicitly, evaluating the cross section perturbations with the help of the SIMULATE-3 system code and the CORE SIM simulator. As an event of interest for the application of the ROM, the instability event that happened in 1996/1997 at the Swedish Power Plant Forsmark-1, was chosen. The ROM input data were adjusted in order to represent the proper operational conditions. As a reference for benchmarking the ROM, the system code output data were used. The time signal for each of the modes were then calculated and some considerations about their stability were made
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