| 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
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 84-89
-
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)
-
Item identification with a space-dependent model of neutron multiplicities and artificial neural networks
- 2023
-
Ingår i: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - 0168-9002. ; 1057
-
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)
-
Development and test of a new verification scheme for transient core simulators
- 2017
-
Ingår i: Transactions of the American Nuclear Society. - 0003-018X. ; 116, s. 1025-1026
-
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)
-
Development of a point-kinetic verification scheme for nuclear reactor applications
- 2017
-
Ingår i: Journal of Computational Physics. - : Elsevier BV. - 1090-2716 .- 0021-9991. ; 339, s. 396-411
-
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)
-
Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
-
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)
-
Estimation of the zero-power reactor transfer from a 3-dimensional core simulator in the frequency domain
- 2016
-
Ingår i: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016.
-
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)
-
Modelling of stationary fluctuations in nuclear reactor cores in the frequency domain
- 2015
-
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
-
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)
-
Description of the models and algorithms used in the coupled CORE SIM neutronic and thermo-hydraulic tool
- 2014
-
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)
-
Development of a fully-consistent Reduced Order Model to study instabilities in Boiling Water Reactors
- 2012
-
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
-
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)
-
Development of a Reduced Order Model and its application to the Forsmark-1 Instability Event of 1996/1997
- 2010
-
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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| 11. |
- Dykin, Victor, 1985, et al.
(författare)
-
Development of a reduced-order model to investigate global/regional/local oscillations in BWRs and study of new stability indicators
- 2012
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report presents the results of some investigations performed at Chalmers University of Technology within the Nordic Thermal-Hydraulic Network (NORTHNET) in the area of stability of Boiling Water Reactors (BWRs). Such systems are known to pos- sibly become unstable under certain conditions, and thus the characterization of their stability properties is of prime importance. Due to the complexity of the problem at hand, a Reduced-Order Model (ROM) was developed, as an alternative to using three- dimensional solvers. The main advantage of using a ROM is that ROMs represent fast running models aimed at catching in a qualitative manner the physical phenomena of im- portance. In addition, the relative simplicity of ROMs compared with three-dimensional solvers leads to the possibility of providing some physical insight into the stability mech- anisms. The ROM developed in this project is unique since it has the ability to model global, regional, and local oscillations, and it is based on four heated channels. The four heated channels are necessary to properly represent the possible excitation of the two first azimuthal modes in case of regional oscillations. A careful examination of the ROM demonstrated that the stability behavior of the system is entirely defined by so-called Cmn-coefficients (assuming that there is no pure density wave oscillation). These coefficients represent the effect of a change of the void fraction on pairs of eigenmodes of the nuclear core. When these coefficients are positive, the system is clearly unstable because of the corresponding positive void feedback. When negative, it was demonstrated, both using the developed ROM and SIMULATE-3K, that the system becomes less stable for Cmn coefficients becoming more negative. A closer examination of the dependence of the Decay Ratio (DR) on the Cmn coefficients using the ROM also demonstrated that for small negative values of the coefficients, a non-monotonic relationship between the DR and the Cmn coefficients exists. Nevertheless, for realistic values of the void reactivity feedback, such a non-monotonic behavior cannot be noticed, because the Cmn coefficients are sufficiently negative. As a consequence, the estimation of the Cmn coefficients opens up the possibility of using such coefficients as a qualitative measure of core stability in a predictive manner. This could be used for instance as a means to compare the relative stability of several core loadings without the need of running lengthy time-dependent three-dimensional core calculations, and could be of great help to nuclear engineers when designing cores.
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| 12. |
- Dykin, Victor, 1985, et al.
(författare)
-
Further Development of the Core Simulator CORE SIM: Extension to coupled capabilities for BWRs
- 2014
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- In nuclear reactors, the monitoring of the nuclear core is of prime importance for guaranteeing the safety of the plant. Assuming stationary conditions, the measurement of process signals using the existing instrumentation supplemented by adequate data acquisition chains allows monitoring fluctuations of the process signals around their mean values. Even though the system does not exhibit any change in the mean values of the process signals, these fluctuations are always present and are the result of e.g. the turbulent character of the cooling flow, coolant evaporation, and/or possible anomalies (excessive vibrations, etc.). These fluctuations (often referred to as “noise”) thus carry some information about the dynamics of the system, and can be used either for core diagnostics/surveillance purposes (i.e. when an anomaly is suspected in the core) or for determining dynamical core parameters/safety coefficients. The main advantage of such techniques relies on the fact that no perturbation of the system is required and that the method is thus a non-intrusive one.The instrumentation present in nuclear core mostly consists of neutron detectors. Many neutron noise diagnostics tasks thus involve an inversion or „unfolding“ procedure, where the neutron noise measured in a few locations throughout the nuclear core is used to determine the root cause (i.e. noise source) responsible for the measured neutron noise. Such an inversion is seldom possible without the knowledge of the so-called reactor transfer function, i.e. the function giving the neutron noise induced by any arbitrary noise source. The Division of Nuclear Engineering at Chalmers University of Technology has been very active for the last ten years in developing computational methods allowing the estimation of such a transfer function for actual reactor cores, i.e. strongly non-homogeneous systems.A numerical tool, named CORE SIM, was developed to estimate the open-loop reactor transfer function. In this tool, the noise source is defined in terms of fluctuations of the macroscopic cross-sections. This report deals with the development of a thermal-hydraulic module coupled to CORE SIM, so that the closed-loop reactor transfer function can also be estimated for Boiling Water Reactors (which is the main type of reactors constituting the Swedish fleet). This module is based on solving the mass, momentum, and enthalpy conservation equations for the fluid, and on solving the heat conduction equation in the solid fuel pellets. Because of the fully coupled neutronic/thermal-hydraulic character of the tool, the noise source can be directly defined in more realistic terms such as perturbations of the flow velocity, temperature, etc. at the inlet of the core. The coupled tool, in addition to be the only one of its kind, has a wide range of applicability.
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| 13. |
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| 14. |
- Dykin, Victor, 1985, et al.
(författare)
-
Investigation of global and regional BWR instabilities with a four heated-channel Reduced Order Model
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 381-400
-
Tidskriftsartikel (refereegranskat)abstract
- The development of an advanced Reduced Order Model (ROM) including four heated channels and meant to study global and regional Boiling Water Reactor (BWR) instabilities is described. The ROM contains three sub-models: a neutron-kinetic model (describing neutron transport), a thermal-hydraulic model (describing fluid transport) 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: the void feedback and the doppler feedback mechanisms. Each of the sub-models is described by a set of reduced ordinary differential equations, derived from the corresponding time- and space-dependent partial differential equations, by using different types of approximations and mathematical techniques that are explained in this paper.One of the novelties of the present ROM is that it takes the effect of the first three neutronic modes into account, namely the fundamental, first, and second azimuthal modes. In order to have a proper representation of both azimuthal modes and of their dependence on the thermal-hydraulic conditions in the heated channels, a four heated channel ROM was constructed. Another novelty of the present work is to develop a special methodology which guarantees the full consistency between the spatial discretization procedures used in the dynamical calculations and the ones implemented in the static case. Accordingly, a re-computation of the static solution based on the CORE SIM tool was embedded into the ROM in such a way that the balance equations expressing the conservation of neutron balance, heat generation, and mass, momentum, enthalpy for the flow, could be fulfilled for the steady-state solution of the coupled neutron-kinetic/thermal-hydraulic problem. Once the static problem is solved, the time-dependent solution in case of a perturbed system can be determined. Moreover, a non-uniform power profile representing different heat production rates in the one- and two-phase regions was introduced into the ROM. Careful attention was paid to the determination of the coupling coefficients for the reactivity effects related to both void fraction and fuel temperature, so that such coefficients correspond to the re-computed static solution. The evaluation of these coefficients was based on the cross-section perturbations estimated by the SIMULATE-3 code, and on the different neutronic eigenmodes of the heterogeneous core determined by the CORE SIM tool.
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| 15. |
- Dykin, Victor, 1985, et al.
(författare)
-
Investigation of local BWR instabilities with a four heated-channel Reduced Order Model
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 53, s. 320-330
-
Tidskriftsartikel (refereegranskat)abstract
- his paper deals with the modeling of Boiling Water Reactor (BWR) local instabilities via so-called Reduced Order Models (ROMs). More specifically, a four-heated channels ROM, which was earlier developed (Dykin et al., submitted for publication), was modified in such a way that the effect of local perturbations could also be accounted for.This model was thereafter used to analyze a local instability event that took place at the Swedish Forsmark-1 BWR in 1996/1997. Such a local instability was driven by unseated fuel assemblies. Comparisons between the results of ROM simulations and actual measurement data demonstrated that the developed ROM was able to correctly reproduce the main features of the event. The ROM has also the ability to give some further physical insights into the phenomena taking place in case of instabilities. For the particular instability event investigated, it was for instance demonstrated that the global and regional oscillation modes were stable, but were excited by the local oscillation acting as an external perturbation. When performing a modal decomposition of the measured neutron flux in case of an instability event driven by a local oscillation, each mode will apparently be excited, whereas in reality such modes might be stable. Such an apparent contradictory behavior is due to the inability of a modal decomposition to catch with only a few modes the spatial dependence of the neutron flux in case of a local oscillation.
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| 16. |
- Dykin, Victor, 1985, et al.
(författare)
-
Investigation of the space-dependent noise induced by propagating density fluctuations
- 2010
-
Ingår i: International Conference on the Physics of Reactors 2010, PHYSOR 2010. - 9781617820014 ; 2, s. 963-978
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The space-dependent behavior of neutron noise induced by a propagating perturbation, represented by the fluctuations of the absorbtion cross section propagating with the coolant of a PWR, is investigated in a one-dimensional one-group approach. The general space-frequency problem is solved for this specific noise source with the help of Greens function technique. All calculations are made in the frame of first-order perturbation theory. The solution is investigated for a different frequencies and system sizes. The limits of point-kinetic and space-dependent behaviour were investigated. An interesting interference phenomenon was found between the point kinetic and the pure space dependent components of the noise for certain combinations of the frequency and system size. The results bear a significance for the dynamics of Molten Salt Reactors (MSR), which will be reported on in a companion paper.
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| 17. |
- Dykin, Victor, 1985
(författare)
-
Noise Applications in Light Water Reactors with Traveling Perturbations
- 2012
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Neutron noise induced by perturbations traveling with the coolant of light water reactors (LWRs) is investigated. Different methods to simulate the effect of propagating perturbations are considered. The studies are performed in both open- and closed-loop systems and summarized in three chapters.In the first chapter, the space-dependence of the neutron noise due to propagating perturbations calculated in one-group theory and one dimension in a pressurized water reactor (PWR) is investigated. A full analytical solution, obtained by the use of Green's function technique, is analyzed for different frequencies and different system sizes. An interesting new interference effect between the point-kinetic and space-dependent components of the induced noise isdiscovered and interpreted in physical terms. A similar investigation is performed in two-group theory for four reactor systems with different neutron spectra. The goal is to investigate the dependence of the properties of the induced neutron noise on the neutron spectrum. The presence of the fluctuations of several cross sections is also analyzed and resulted in qualitatively and quantitatively new characteristics of the induced noise. Further, a simple numerical Monte Carlo-based model to simulate the boiling process in a boiling water reactor (BWR) heated channel, is constructed. The output of the model is then used to estimate the local component of the neutron noise induced by density fluctuations in the coolant numerically convoluting it with proper transfer functions.In the second chapter, a four-heated channel reduced order model (ROM), accounting for the first three neutronic modes, is constructed to study both global and regional instabilities. Some additional modifications compared with the earlier-developed models are performed to improve the consistency of the model. It is shown that the ROM is capable to reproduce the main features of core-wide instabilities. Moreover, it is proven that the inclusion of both azimuthal modes brings some importance for the correct identification of stability boundaries. The ROM is also extended to simulate the effect of local instabilities, such as the Forsmark-1 instability event of 1996/1997. A good qualitative agreement with real measurements is found.In the last chapter, a number of the applications of the noise diagnostics based on the foregoing calculations are discussed. The case when the neutronic response of the reactor is affected bya non-white driving force (propagating perturbation) is studied. It is also investigated how the accuracy of the determination of the so-called decay ratio (DR) of the system, based on the assumption of a white noise driving force, deteriorates with deviations from the white noise character of the driving force. Furthermore, the earlier developed ROM is applied to analyze what stability indicators other than the DR can be used to describe the stability of the system. As a candidate, the coupling reactivity coefficients are chosen and their dependence on the DR is investigated. It is shown that such a dependence deviates form the conventional one, presumably caused by the inherent inertia of the system. Finally, two techniques, one based on the break-frequency of auto power spectral density (APSD) of the neutron noise and another on the transit times of propagating void fluctuations are discussed for reconstructing the axial void profile from the Monte-Carlo simulated neutron noise. It is shown that both methods provide promising results.
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| 18. |
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| 19. |
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| 20. |
- Dykin, Victor, 1985, et al.
(författare)
-
Predictive BWR core stability using feedback reactivity coefficients projected on neutronic eigenmodes
- 2019
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 124, s. 1-8
-
Tidskriftsartikel (refereegranskat)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.
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| 21. |
- Dykin, Victor, 1985, et al.
(författare)
-
Qualitative and quantitative investigation of the propagation noise in various reactor systems
- 2014
-
Ingår i: Progress in Nuclear Energy. - : Elsevier BV. - 0149-1970. ; 70, s. 98-111
-
Tidskriftsartikel (refereegranskat)abstract
- The space-dependent neutron noise, induced by propagating perturbations (propagation noise for short) is investigated in a one-dimensional homogeneous model of various reactor systems. By using two-group theory, the noise in both the fast and the thermal group is calculated. The purpose is to investigate the dependence of the properties of the space-dependent fast and thermal propagation noise on the static neutron spectrum as well as on the presence of the fluctuations of several cross sections. The motivation for this study arose in connection with recent work on neutron noise in molten salt reactors (MSR) with propagating fuel of various compositions. Some new features of the induced noise were observed, but it was not clear whether these were due to the propagating perturbation alone, or to the propagation of the fuel and hence that of the delayed neutron precursors. The present study serves to clarify the significance of the spectral properties of the different cores through calculating the propagation noise in four different reactor systems, as well as considering the influence of the perturbation of the various cross sections. By comparing the results with those obtained in MSR, the effect of the moving fuel on the propagation noise is clarified. It is shown that in fast systems the noise in the fast group is much larger than in the thermal group and hence can gain diagnostic importance. It is also shown that the coexistence of several cross section fluctuations leads to qualitatively and quantitatively new characteristics of the noise, hence it is important to model the effect of e.g. temperature fluctuations of the coolant in a proper way. (C) 2013 Elsevier Ltd. All rights reserved.
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| 22. |
- Dykin, Victor, 1985, et al.
(författare)
-
Remark on the neutron noise by propagating perturbations in a MSR
- 2014
-
Ingår i: Proceedings of the International Conference on Physics of Reactors, PHYSOR 2014.
-
Konferensbidrag (refereegranskat)abstract
- The neutron noise induced by propagating perturbations in a bare 1-D Molten Salt Reactor (MSR) model is calculated and analyzed using one-group diffusion theory. The neutron noise for different noise sources of which two have not been accounted for, corresponding to the fluctuations of the fission and absorption cross sections as well as to the fuel velocity is calculated and the results are qualitatively compared. Unlike in previous work, the solution is obtained through the matrix Green's function of the flux and precursor equations being kept separate. It is shown that in the case when the noise is represented by the fluctuations of the fission cross-section, the noise source attains a complex structure which is different from that in traditional reactors. On the other hand, in the cases investigated, despite all qualitative differences in the noise calculation procedure as well as in the structure of the noise source, it turns out that the noise induced by the absorption and the fission cross sections follow a similar behaviour. In addition, it is observed that the inclusion of the fluctuations in the fuel velocity examined in this paper slightly suppresses the total neutron noise for low frequency region i.e. below ∼ 2 Hz but on the other hand it enhances the latter one by one of order of magnitude for high frequencies i.e. above ∼ 2 Hz compared to the effect of other noise sources. The results contribute to the understanding and interpretation of the neutron noise in MSRs.
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| 23. |
- Dykin, Victor, 1985, et al.
(författare)
-
Remark on the neutron noise induced by propagating perturbations in an MSR
- 2016
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 90, s. 93-105
-
Tidskriftsartikel (refereegranskat)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.
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| 24. |
- Dykin, Victor, 1985, et al.
(författare)
-
Remark on the role of the driving force in BWR instability
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1544-1552
-
Tidskriftsartikel (refereegranskat)abstract
- Simple models of BWR instability, used e.g. in understanding the role of the various oscillation modes inthe overall stability of the plant, assume that each oscillation mode can be described by a second ordersystem (a damped harmonic oscillator) driven by a white noise driving force. Change of the decay ratio(DR) of the observed signal is, as a rule, associated with the changing of the parameters of the dampedoscillator, mainly its damping coefficient, and is interpreted in terms of the change of the stability ofthe system. However, conceptually, one cannot exclude cases when the change of the response of a drivendamped oscillator is due to the change of the properties of the driving force. In this work we investigatethe effect of a non-white driving force on the behaviour of the system. A question of interest is howchanges of the spectrum of the driving force influence the observed autocorrelation function (ACF) ofthe resulting signal. Hence we calculate the response of a damped harmonic oscillator driven by anon-white driving force, corresponding to the reactivity effect of propagating density fluctuations intwo-phase flow. It is shown how in some special cases such a driving force, when interpreting the neutronnoise as if induced by a white noise driving source, can lead to an erroneous conclusion regarding thestability of the system. It is also concluded that in the practically interesting cases the effect of the coloureddriving force, arising from propagating density fluctuations, is negligible.
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| 25. |
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| 26. |
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| 27. |
- Dykin, Victor, 1985, et al.
(författare)
-
Ringhals Diagnostics and Monitoring, Annual Research Report 2015
- 2015
-
Rapport (övrigt vetenskapligt/konstnärligt)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.
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| 28. |
- Dykin, Victor, 1985, et al.
(författare)
-
Ringhals Diagnostics and Monitoring, Final Research Report 2012-2014
- 2014
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report gives an account of the work performed by the Department of Nuclear Engineering, Chalmers, in the frame of research collaboration with Ringhals, Vattenfall AB, contract No. 630217-031. The contract constitutes a 3-year co-operative research work concerning diagnostics and monitoring of the BWR and PWR units. The work in thecontract has been performed between January 1st 2012, and December 31st, 2014. During this period, we have worked with four main items as follows:1. Development and application of the analysis method of core barrel vibrations, developed in the previous Stages, to three ex-core measurements performed during several cycles in R2, R3 and R4. What regards R2, this was the first attempt to analyze ex-core measurements taken at BOC, MOC and EOC, with the new curve-fitting procedure;2. Investigation of the ultra-low frequency oscillations in reactor power in R4;3. Development of the theory and simulations in order to determine the void content in R1 from the analysis of in-core measurements;4. Evaluation of the measurements made in R1 with the use of 4 LPRMs and one TIP detector, for testing the velocity and void fraction profile reconstruction methods.This work was performed at the Department of Nuclear Engineering, Chalmers University of Technology by Victor Dykin, Cristina Montalvo (visitor from the TechnicalUniversity of Madrid), Imre Pázsit (project co-ordinator) and Henrik Nylén, who was also the contact person at Ringhals.
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| 29. |
- Dykin, Victor, 1985, et al.
(författare)
-
Simulation of in-core neutron noise measurements for axial void profile reconstruction in boiling water reactors
- 2012
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 183:3, s. 354-366
-
Konferensbidrag (refereegranskat)abstract
- This paper reports on the development and application of a method of emulating bubbly flow by generating bubbles with random sampling methods. The purpose of the modeling is that by using the simulated random two phase flow as input, one can generate "synthetic" neutron noise signals by convoluting the input with a simplified neuronic transfer function, on which the possibility of reconstructing the axial void profile from in-core neutron noise measurements can be studied by standard spectral noise analysis methods. The long term goal of this work is to elaborate methods of neutron noise analysis, by which the local void fraction in a boiling water reactor can be determined by measurements. In this preliminary stage, two methods for the reconstruction of the axial void and the velocity profiles are discussed. The first method is based on the break frequency of the neutron auto-power spectrum, whereas the second method only utilizes the information in the transit time of the void fluctuations between axial pairs of neutron detectors. A clear and monotonic relationship between the chosen observables and the two-phase flow properties was found, but an accurate determination of the void fraction requires further development and testing of the various unfolding alternatives.
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| 30. |
- Dykin, Victor, 1985
(författare)
-
The Effect of Different Perturbations on the Stability Analysis of Light Water Reactors
- 2010
-
Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Neutron noise analysis techniques are studied and developed, with primary useof determining the stability of Boiling Water Reactors (BWRs). In particular, the role ofa specific perturbation prevailing in Light Water Reactors, the propagating densityperturbation, in the stability of BWRs and on the noise field of LWRs in general, isinvestigated by considering three topics.In the first topics, we investigate how the neutronic response of the reactor, usuallydescribed as a second order system driven by a white noise driving force, is affected bya non-white driving force. This latter arises from the reactivity effect of the propagatingdensity perturbations. The investigation is performed by using spectral and correlationanalysis. Propagating perturbations with different velocities are analyzed. We investigatehow the accuracy of the determination of the so-called decay ratio (DR) of the system,based on the assumption of white noise driving force, deteriorateswith deviations from the white noise character of the driving force.In the second topics, the space dependence of the neutron noise, induced bypropagating density perturbations, represented through the perturbation ofthe absorption, is determined and discussed. A full analytical solution was obtainedby the use of the Green's function technique. The solution was analyzed for differentfrequencies and different system sizes. An interesting new interference effectbetween the point-kinetic and space-dependent components of the induced noise wasdiscovered and interpreted in physical terms.In the last topics, a non-linear stability analysis of a BWR is performed,using so called Reduced Order Model (ROM) techniques. A ROM is usually constructedby reducing the full set of 3D space-time dependent neutron-kinetics,thermal-hydraulics and heat transfer equations to time-dependent ones, byconsidering space dependence in a lumped parameter model (one or two discrete channels).The main novelty of our work is to treat thespace dependence by four heated channels. This extension makes itpossible to account for the effect of three neutronic modes: fundamental, firstand second azimuthal ones. The Forsmark-1 instability event in 1996/1997was chosen to be investigated by the ROM developed in this work. The reactor response was determinedfor various operational points to identify the stable/unstable reactorbehavior. The suitability of using the DR as the stability parameter in case of non-linear oscillationsis also being investigated.
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| 31. |
|
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| 32. |
- Dykin, Victor, 1985, et al.
(författare)
-
The Molten Salt Reactor Point-Kinetic Component of Neutron Noise in Two-Group Diffusion Theory
- 2016
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 193:3, s. 404-415
-
Tidskriftsartikel (refereegranskat)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.
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| 33. |
- Dykin, Victor, 1985, et al.
(författare)
-
The two-group point-kinetic component of neutron noise in a MSR
- 2014
-
Ingår i: Proceedings of the International Conference on Physics of Reactors, PHYSOR 2014.
-
Konferensbidrag (refereegranskat)abstract
- The calculation of the point kinetic component of the neutron noise in two-group diffusion theory in Molten Salt Reactors (MSRs) using different techniques is discussed. First, the point kinetic component of the noise is calculated from the full space-frequency dependent solution analytically by a projection to the static adjoint. Then, the point-kinetic solution is derived by solving the simplified point kinetic equations. Both results are thereafter analyzed and compared quantitatively. This comparison shows that the solution of the simplified point kinetic equations significantly differs from the exact one and cannot reconstruct some important features of the true solution. The similar discrepancies between two methods were also observed and confirmed in earlier one-group MSR calculations.
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| 34. |
- Dykin, Victor, 1985, et al.
(författare)
-
User's manual of the coupled CORE SIM neutronic and thermo-hydraulic tool
- 2014
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report presents how to use the coupled CORE SIM neutronic/thermo-hydraulic tool. The models and algorithms used in the coupled version of CORE SIM, as well as the demonstration of the tool were already presented in two companion reports. 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 entirely MATLAB 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 Boiling Water Reactors (PWR and BWR,respectively). The tool is freely available on direct request to the authors of the present report.
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| 35. |
- Dykin, Victor, 1985
(författare)
-
Using neutron noise to determine void fraction
- 2013
-
Ingår i: Nuclear Engineering International. - 0029-5507. ; 58:709, s. 16-17
-
Tidskriftsartikel (refereegranskat)abstract
- A long-known dependence of neutron noise phenomena on void fraction has become the basis of a useful online monitoring technique that backs up computational models and improves operations. Calculations have shown good agreement with simulated data; calculations with real data are expected later this year.
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| 36. |
- Dykin, Victor, 1985, et al.
(författare)
-
Void reactivity (Cmn) coefficients as indicators of boiling water reactor stability
- 2016
-
Ingår i: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016. - 9781510825734 ; 6, s. 3571-3578
-
Konferensbidrag (refereegranskat)abstract
- This paper presents the results of some investigations performed in the area of the stability of Boiling Water Reactors (BWRs). These systems are known to become unstable under certain conditions, and thus the development of different methods for the quantification of their stability properties is of prime importance. Preliminary studies with Reduced Order Models showed that the stability behavior of the system can be described in terms of so-called Cmn-coefficients. These coefficients represent the effect of a change in the void fraction on the eigenmodes of a nuclear core. It turned out that when these coefficients are positive, the system is clearly unstable because of the corresponding positive void feedback. On the other hand, in the case of negative Cmn-coefficients, it has been found, both using ROMs and dynamic core simulators like SIMULATE-3K, that the system becomes less stable for the Cmn coefficients becoming more negative. A thorough examination of the dependence of Decay Ratio on the Cmn coefficients using ROMs also demonstrated that there is a strong correlation between the DR and the Cmn coefficients what regards the stability properties of a BWR system. As a result, the estimation of the Cmn coefficients opens up the possibility of using such coefficients as a qualitative measure of core stability in a predictive manner. This could be used for instance as means to compare the relative stability of several core loadings without the need of running lengthy time-dependent three-dimensional core calculations, and could be of great help to nuclear engineers when designing nuclear cores. Therefore the main purpose of the paper is to investigate the possibility of using Cmn-coefficients as an alternative stability indicator for Boiling Water Reactors, as well as to determine their advantages and disadvantages as compared to the traditional Decay Ratio.
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| 37. |
- Hursin, Mathieu, et al.
(författare)
-
Measurement of the gas velocity in a water-air mixture in crocus by neutron noise technique
- 2019
-
Ingår i: International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019. ; , s. 2696-2703
-
Konferensbidrag (refereegranskat)abstract
- © 2019 American Nuclear Society. All rights reserved. The possibility to measure the gas phase velocity in a two-component mixture with neutron noise techniques is demonstrated in the zero-power reactor CROCUS of the Ecole Polytechnique Federale de Lausanne. It is the first step toward the experimental validation of a theoretical method aiming at the reconstruction of the void profile in a BWR channel. For this experiment, a channel is installed in the water reflector of CROCUS and two-component mixtures are generated inside the channel through injection of air at various flow rates. The signal fluctuations of two neutron detectors located at different elevations next to the channel are recorded and their Cross Power Spectral Density analyzed with various techniques to determine the transit time of the gas phase and its velocity. Experimental results are compared with predictions obtained with the TRACE thermal-hydraulic code. Results disagree in their magnitudes but the evolution of the gas velocity with the air injection rate are similar.
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| 38. |
- Hursin, Mathieu, et al.
(författare)
-
Measurement of the Gas Velocity in a Water-Air Mixture in CROCUS Using Neutron Noise Techniques
- 2020
-
Ingår i: Nuclear Technology. - : Informa UK Limited. - 0029-5450 .- 1943-7471. ; 206:10, s. 1566-1583
-
Tidskriftsartikel (refereegranskat)abstract
- The possibility of measuring the gas-phase velocity in a two-phase mixture through the use of neutron noise techniques is demonstrated in the zero-power reactor CROCUS of the Ecole Polytechnique Federale de Lausanne. It is the first step toward the experimental validation of an existing theoretical model whose objective is the reconstruction of the void profile in a channel. The use of zero-power research reactors is advantageous due to their clean environment in terms of signal fluctuations. To this end, a channel was installed in the reflector of CROCUS. A two-component mixture is generated inside the channel through the injection of compressed air. The signal fluctuations of neutron detectors located at various axial locations next to the channel are processed to determine the transit time of the gas phase between detectors. Four methods are presented based on the detector signal time series either in the time domain (time correlations between signals) or in the frequency domain (phase of the cross-power spectral density. All four methods returned consistent transit times and similar experimental uncertainty. The largest possible gas injection rates as well as the highest possible neutron flux level improve the visibility of the traveling perturbation and reduce the experimental uncertainty on the transit time for a given acquisition time. © 2020, © 2020 American Nuclear Society.
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| 39. |
- Hursin, Mathieu, et al.
(författare)
-
Validation of axial void profile measured by neutron noise techniques in crocus
- 2020
-
Ingår i: International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020. - : EDP Sciences. ; 2020-March, s. 1586-1593
-
Konferensbidrag (refereegranskat)abstract
- Recently a joint project has been carried out between the Paul Scherrer Institut, the Ecole Polytechnique Federale de Lausanne and swissnuclear, an industrial partner, in order to determine the axial void distribution in a channel installed in the reflector of the zero power research reactor CROCUS, using neutron noise techniques. The main objective of the present paper is to report on the validation of the results against an alternative measurement technique using gamma-ray attenuation and simulations with the TRACE code. For the gamma-ray attenuation experiments, the channel used in CROCUS is installed out of the core in a Plexiglass water tank. The source and detector are fixed and the channel is moved axially to keep the geometry of the source/detector arrangement untouched. This is key to measure the void effect by gamma attenuation due to the low contrast of this technique. The paper compares the experimental results obtained with both techniques, with the outcomes of simulations carried out with the TRACE code. Even though the quantitative void fraction estimations are not consistent, the trends obtained with the simulation and experimental techniques are the same. The discrepancies between the various experimental techniques and the simulation outcomes are related to the heterogeneous distribution of the water-air mixture in the radial sections of the channel.
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| 40. |
- Jonsson, Anders, 1984, et al.
(författare)
-
Analytical investigation of the properties of the neutron noise induced by vibrating absorber and fuel rods
- 2012
-
Ingår i: Kerntechnik. - : Walter de Gruyter GmbH. - 0932-3902 .- 2195-8580. ; 77:5, s. 371-380
-
Tidskriftsartikel (refereegranskat)abstract
- Analytical solution methods for the neutron noise in a one-dimensional multi-region system in two-group theory, which have so far been based on the adjoint function technique, are extended here to using the forward Green's function technique. The forward Green's functions were calculated analytically for a noise source in a core surrounded by reflector regions at both sides. It is shown that with symbolic computation methods, the forward Green's function can be used for the calculation of the space- and frequency-dependent noise in the first order approximation for arbitrary noise sources which have an analytical representation. The properties of the induced neutron noise were investigated for vibrations of both absorbers and fuel assemblies, with two representations of the noise sources: a point-like source which corresponds to the vibrations of a fuel rod, and a finite width source which corresponds to vibrations of a fuel assembly The contributions of the components induced by the fluctuations of the various types of macroscopic cross sections in the total noise are also discussed and the information content of the noise in the fast group is explored for the identification of fuel assembly vibrations.
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| 41. |
- Lange, Carsten, et al.
(författare)
-
Comments on local power oscillation phenomenon at BWRs
- 2012
-
Ingår i: Progress in Nuclear Energy. - : Elsevier BV. - 0149-1970. ; 60, s. 73-88
-
Tidskriftsartikel (refereegranskat)abstract
- Under the framework of BWR stability analysis, local neutron-flux oscillation events have attracted the attention of a number of researchers. In 1996, an unusual instability event occurred at Forsmark-1 in which an irregular oscillation pattern with highly localized, relatively large-amplitude oscillations were measured. Some authors assumed that this behaviour was caused by the superposition of stable spatial mode limit cycle oscillations, where the BWR core as a neutron kinetics/-thermal-hydraulic coupled system is unstable. Subsequent time-series analysis of the local power range monitor (LPRM) signals resulted in a space-dependent decay ratio, an inexplicable result. Furthermore, noise analysis-based localization techniques pointed towards the existence of two strong “perturbation sources” in one of the two halves of the core, one of them coinciding with the radial position of an unseated bundle. In the scope of theoretical work, the possibility of a space-dependent decay ratio was discussed but not comprehensively understood. Motivated by these findings, the effect of local neutron-flux oscillations on the stability behaviour of BWR is discussed, and one possible interpretation is proposed which is able to explain the space-dependent decay ratio as well as the long term oscillation pattern. The RAM–ROM method is applied to a Forsmark measurement case, where an irregular oscillation pattern was found and to an operational point (KKB-B8) of NPP Brunsbüttel, where a local neutron-flux oscillation is superimposed on an unstable global power oscillation. The effect of the local neutron flux oscillating sources on the space- and time-dependent neutron field is described by a rigorous application of the mode expansion approach. The consequences to signal analysis are then discussed. It will be pointed out in the paper that when a BWR system is stable with regards to power oscillations but is driven by local neutron-flux oscillating sources, the decay ratio does not indicate the real BWR stability behaviour.
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| 42. |
- Montalvo, C., et al.
(författare)
-
First evidence of the pivotal motion ("tilting mode") of the core Barrel in the RINGHALS-4 PWR
- 2016
-
Ingår i: Physics of Reactors 2016, PHYSOR 2016: Unifying Theory and Experiments in the 21st Century. - 9781510825734 ; 4, s. 2571-2579
-
Konferensbidrag (refereegranskat)abstract
- The Division of Subatomic Physics and Plasma Physics (formerly Division of Nuclear Engineering) in Chalmers, Göteborg, and the Ringhals Nuclear Plant have investigated the core barrel vibrations in the Ringhals PWRs over the last 20 years. Based on the different symmetry properties of the vibration modes, a mode separation technique was developed to enhance the contributions from the different modes. Recent observations of wear at both the lower and upper core-barrel-support structures in the Ringhals PWRs indicated that vibration modes of the core barrel other than pendular (beam mode) and shell mode are likely to occur. A beam mode type movement alone is not able to explain such a wear, and therefore, it is fair to assume that the vibration mode in question is a small amplitude periodic tilting movement of the core barrel around a horizontal, diagonal pivot at the half height of the core. In this work, ex-core data taken in the Ringhals-4 PWR were analyzed in order to find evidence of such a tilting movement. First, cross spectra between various ex-core detectors were calculated and analyzed to locate the frequency range of the new vibrational mode. Then, a model based on symmetry considerations was derived in order to extract the sought mode from the spectra. The measurements were evaluated by the new mode enhancement technique. The results show that it is possible to enhance such a mode and find it in the spectra by properly combining the signals in the time domain.
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| 43. |
- Pazsit, Imre, 1948, et al.
(författare)
-
A possible application of catastrophe theory to boiling water reactor instability
- 2020
-
Ingår i: Progress in Nuclear Energy. - : Elsevier BV. - 0149-1970. ; 118
-
Tidskriftsartikel (refereegranskat)abstract
- It is known that the stability of boiling water reactors (BWRs), when quantified with the so-called decay ratio as the stability parameter, may show seemingly abrupt changes, despite the smooth variations of the control parameters (reactor power and core flow). There is also evidence of the fact that the stability properties can exhibit a hysteresis effect when moving back and forth on the same path on the power-flow map. The most common explanation of the abrupt change is based on the co-existence of two different types of instabilities (global and regional), and their interplay (van der Hagen et al., 1994; Pázsit, 1995). In this paper we suggest an alternative phenomenological explanation, which only assumes the existence of one single mode of instability. We propose the hypothesis that the decay ratio of one single mode of a complex, many-variable non-linear system might obey a cusp catastrophe as a function of the control parameters. Such a phenomenological model can explain both the discontinuous variation of the decay ratio, as well as the hysteresis effect. Since a cusp-type behaviour implies that the decay ratio is many-valued in a certain region of the power-flow map, a mechanism is suggested how a Hopf bifurcation with multiplicative noise can lead to such a behaviour.
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| 44. |
- Pazsit, Imre, 1948, et al.
(författare)
-
Development of a new method to determine the axial void velocity profile in BWRs from measurements of the in-core neutron noise
- 2021
-
Ingår i: Progress in Nuclear Energy. - : Elsevier BV. - 0149-1970. ; 138
-
Tidskriftsartikel (refereegranskat)abstract
- Determination of the local void fraction in BWRs from in-core neutron noise measurements requires the knowledge of the axial velocity of the void. The purpose of this paper is to revisit the problem of determining the axial void velocity profile from the transit times of the void between axially placed detectors, determined from in-core neutron noise measurements. In order to determine a realistic velocity profile which shows an inflection point and hence has to be at least a third order polynomial, one needs four transit times and hence five in-core detectors at various axial elevations, whereas the standard instrumentation usually consists only of four in-core detectors. Attempts to determine a fourth transit time by adding a TIP detector to the existing four LPRMs and cross-correlate it with any of the LPRMs have been unsuccessful so far. In this paper we thus propose another approach, where the TIP detector is only used for the determination of the axial position of the onset of boiling. By this approach it is sufficient to use only three transit times. Moreover, with another parametrisation of the velocity profile, it is possible to reconstruct the velocity profile even without knowing the onset point of boiling, in which case the TIP is not needed, although at the expense of a less flexible modelling of the velocity profile. In the paper the principles are presented, and the strategy is demonstrated by concrete examples, with a comparison of the performance of the two different ways of modelling the velocity profile. The method is tested also on velocity profiles supplied by system codes, as well as on transit times from neutron noise measurements.
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| 45. |
|
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| 46. |
|
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| 47. |
- Pazsit, Imre, 1948, et al.
(författare)
-
Investigation of the space-dependent noise induced by propagating perturbations
- 2010
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:10, s. 1329-1340
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Tidskriftsartikel (refereegranskat)abstract
- The space-dependent behaviour of the neutron noise, induced by perturbations represented by the fluctuations of the absorption cross sections, propagating with the coolant of a PWR, is investigated in a one-dimensional one-group approach. The general space–frequency dependent problem is solved for this specific noise source with the help of the Green’s function technique. All calculations are made in the frame of first-order perturbation theory. The solution is investigated for different frequencies and system sizes. The limits of point kinetic and space-dependent behaviour were investigated. An interesting interference phenomenon was found between the point kinetic and the pure space dependent components of the noise for certain domains of the frequency and system size. The results bear some significance for the dynamics of Molten Salt Reactors (MSR), which is reported on in a companion paper.
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| 48. |
- Pazsit, Imre, 1948, et al.
(författare)
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Kinetics, dynamics, and neutron noise in stationary MSRs
- 2017
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Ingår i: Molten Salt Reactors and Thorium Energy. ; , s. 111-166
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Bokkapitel (övrigt vetenskapligt/konstnärligt)abstract
- This chapter discusses the statics, kinetics, and dynamics of molten salt reactors in a simple model that allows analytical solutions. Due to this, and the introduction of some further limiting cases, the chapter offers substantial insight into and understanding of the physics and the neutronic behavior of molten salt systems with circulating fuel. After a discussion of the properties of the model and the limiting cases for the static equations, a treatment of space-time transients is introduced, and some limiting cases are solved explicitly. Thereafter the kinetic and dynamic response of the reactor is derived by calculating the space-frequency-dependent neutron fluctuations (neutron noise) induced by small stationary perturbations of the system parameters. The validity of the point kinetic approximation and the calculation of the point kinetic component of the noise are discussed. Finally, the neutron noise induced by various perturbations propagating with the circulating fuel is calculated and discussed.
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| 49. |
- Pazsit, Imre, 1948, et al.
(författare)
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Reconstructing the axial void velocity profile in BWRS from measurements of the in-core neutron noise
- 2020
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Ingår i: International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020. - : EDP Sciences. ; 2020-March, s. 211-218
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Konferensbidrag (refereegranskat)abstract
- The problem of determining the axial velocity profile from the in-core neutron noise measurements is revisited, with the purpose of developing an objective method for the determination of the void fraction. Until now it was assumed that in order to determine a realistic velocity profile which shows an inflection point and hence has to be at least a third order polynomial, one needs four transit times and hence five in-core detectors at various axial elevations. However, attempts to determine a fourth transit time by adding a TIP detector to the existing four LPRMs and cross-correlate it with any of the LPRMs were unsuccessful so far. In this paper we thus propose another approach, where the TIP detector is only used for the determination of the axial position of the onset of boiling. By this approach it is sufficient to use only three transit times. Moreover, with another parametrisation of the velocity profile, it is possible to reconstruct the velocity profile even without knowing the onset point of boiling, in which case the TIP is not needed. In the paper the principles are explained and the strategy is demonstrated by concrete examples.
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| 50. |
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