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Sökning: L773:0306 4549 OR L773:1873 2100

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51.
  • Dufek, Jan, 1978 (författare)
  • Building the nodal nuclear data dependences in a many-dimensional state-variable space
  • 2011
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:7, s. 1569-1577
  • Tidskriftsartikel (refereegranskat)abstract
    • We present new methods for building the polynomial-regression based nodal nuclear data models. Thedata models can reflect dependences on a large number of state variables, and they can consider varioushistory effects. Suitable multivariate polynomials that approximate the nodal data dependences are identifiedefficiently in an iterative manner. The history effects are analysed using a new sampling scheme forlattice calculations where the traditional base burnup and branch calculations are replaced by a largenumber of diverse burnup histories. The total number of lattice calculations is controlled so that the datamodels are built to a required accuracy.
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52.
  • Dufek, Jan, et al. (författare)
  • Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 94, s. 415-421
  • Tidskriftsartikel (refereegranskat)abstract
    • Previous studies raised a question about the level of a possible correlation of errors in the cumulative Monte Carlo fission source and the fundamental-mode eigenvector of the fission matrix. A number of new methods tally the fission matrix during the actual Monte Carlo criticality calculation, and use its fundamental-mode eigenvector for various tasks. The methods assume the fission matrix eigenvector is a better representation of the fission source distribution than the actual Monte Carlo fission source, although the fission matrix and its eigenvectors do contain statistical and other errors. A recent study showed that the eigenvector could be used for an unbiased estimation of errors in the cumulative fission source if the errors in the eigenvector and the cumulative fission source were not correlated. Here we present new numerical study results that answer the question about the level of the possible error correlation. The results may be of importance to all methods that use the fission matrix. New numerical tests show that the error correlation is present at a level which strongly depends on properties of the spatial mesh used for tallying the fission matrix. The error correlation is relatively strong when the mesh is coarse, while the correlation weakens as the mesh gets finer. We suggest that the coarseness of the mesh is measured in terms of the value of the largest element in the tallied fission matrix as that way accounts for the mesh as well as system properties. In our test simulations, we observe only negligible error correlations when the value of the largest element in the fission matrix is about 0.1. Relatively strong error correlations appear when the value of the largest element in the fission matrix raises above about 0.5. We also study the effect of the error correlations on accuracy of the eigenvector-based error estimator. The numerical tests show that the eigenvector-based estimator consistently underestimates the errors in the cumulative fission source when a strong correlation is present between the errors in the fission matrix eigenvector and the cumulative fission source (i.e., when the mesh is too coarse). The error estimates are distributed around the real error value when the mesh is sufficiently fine.
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53.
  • Dufek, Jan, et al. (författare)
  • Derivation of a stable coupling scheme for Monte Carlo burnup calculations with the thermal-hydraulic feedback
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 62, s. 260-263
  • Tidskriftsartikel (refereegranskat)abstract
    • Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydraulic conditions) is calculated iteratively by the stochastic approximation; the fuel nuclide densities and thermal-hydraulic conditions are iterated simultaneously. This coupling scheme is derived as stable in theory; i.e.; its stability is not conditioned by the choice of time steps.
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54.
  • Dufek, Jan, et al. (författare)
  • Description of a stable scheme for steady-state coupled Monte Carlo-thermal-hydraulic calculations
  • 2014
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 68, s. 1-3
  • Tidskriftsartikel (refereegranskat)abstract
    • We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal-hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal-hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us, we have decided to make it available via this note.
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55.
  • Dufek, Jan, et al. (författare)
  • Fission matrix based Monte Carlo criticality calculations
  • 2009
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1270-1275
  • Tidskriftsartikel (refereegranskat)abstract
    • We have described a fission matrix based method that allows to cancel the inactive cycles in Monte Carlo criticality calculations. The fission matrix must be sampled in the course of the Monte Carlo calculation using a space mesh with sufficiently small zones as it causes the fission matrix be insensitive to errors in the initial fission source. The k(eff) and other quantities can be derived by means of the final fission matrix. The confidence interval for the k(eff) estimate can be conservatively determined via the variance in the fission matrix.
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56.
  • Dufek, Jan, 1978-, et al. (författare)
  • Monte Carlo criticality calculations accelerated by a growing neutron population
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 94, s. 16-21
  • Tidskriftsartikel (refereegranskat)abstract
    • We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. As the efficiency of Monte Carlo criticality simulations is sensitive to the selected neutron population size, the method attempts to achieve the acceleration via on-the-fly control of the neutron population size. The neutron population size is gradually increased over successive criticality cycles so that the fission source bias amounts to a specific fraction of the total error in the cumulative fission source. An optimal setting then gives a reasonably small neutron population size, allowing for an efficient source iteration; at the same time the neutron population size is chosen large enough to ensure a sufficiently small source bias, such that does not limit accuracy of the simulation.
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57.
  • Dufek, Jan, et al. (författare)
  • Numerical stability of the predictor-corrector method in Monte Carlo burnup calculations of critical reactors
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 56, s. 34-38
  • Tidskriftsartikel (refereegranskat)abstract
    • Monte Carlo burnup codes use various schemes to solve the coupled criticality and burnup equations. Previous studies have shown that the simplest methods, such as the beginning-of-step and middle-of-step constant flux approximations, are numerically unstable in fuel cycle calculations of critical reactors. Here we show that even the predictor-corrector methods that are implemented in established Monte Carlo burnup codes can be numerically unstable in cycle calculations of large systems.
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58.
  • Dufek, Jan, 1978-, et al. (författare)
  • Optimal time step length and statistics in Monte Carlo burnup simulations
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 139
  • Tidskriftsartikel (refereegranskat)abstract
    • Monte Carlo burnup simulations continue to be seen as computationally very expensive numerical routines despite recent developments of associated methods. Here, we suggest a way of improving the computing efficiency via optimisation of the length of the time steps and the number of neutron histories that are simulated at each Monte Carlo criticality run. So far, users of Monte Carlo burnup codes have been required to set these parameters at will; however, an inadequate choice of these free parameters can severely worsen the computing efficiency. We have tested a large number of combinations of the free parameters on a simplified and fast solver, and we have observed that the computing efficiency was maximized when the computing cost of all Monte Carlo neutron transport calculations (summed over all time steps) was approximately comparable to costs of other procedures (all depletion simulations, the loading and processing of neutron cross sections, etc.). In this technical note, we demonstrate these results, and we also derive a simple theoretical model of the convergence of Monte Carlo burnup simulations that conforms to these numerical results. Here, we also suggest a straightforward way to automatise the selection of the optimal values of the free parameters for Monte Carlo burnup simulations.
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59.
  • Dufek, Jan, et al. (författare)
  • Stability and convergence problems of the Monte Carlo fission matrix acceleration methods
  • 2009
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1648-1651
  • Tidskriftsartikel (refereegranskat)abstract
    • The Monte Carlo fission matrix acceleration methods aim at accelerating the convergence of the fission source in inactive cycles of Monte Carlo criticality calculations. In practice, however, these methods may corrupt the fission source, or slow down its convergence. These phenomena have not been completely understood so far. We demonstrate the convergence problems, and explain their reasons.
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60.
  • Dufek, Jan, et al. (författare)
  • The stochastic implicit Euler method - A stable coupling scheme for Monte Carlo burnup calculations
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 60, s. 295-300
  • Tidskriftsartikel (refereegranskat)abstract
    • Existing Monte Carlo burnup codes use various schemes to solve the coupled criticality and bumup equations. Previous studies have shown that the coupling schemes of the existing Monte Carlo burnup codes can be numerically unstable. Here we develop the Stochastic Implicit Euler method - a stable and efficient new coupling scheme. The implicit solution is obtained by the stochastic approximation at each time step. Our test calculations demonstrate that the Stochastic Implicit Euler method can provide an accurate solution to problems where the methods in the existing Monte Carlo burnup codes fail.
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61.
  • Dufek, Jan, et al. (författare)
  • Time step length versus efficiency of Monte Carlo burnup calculations
  • 2014
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 409-412
  • Tidskriftsartikel (refereegranskat)abstract
    • We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy.
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62.
  • 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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63.
  • 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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64.
  • 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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65.
  • 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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66.
  • 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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67.
  • Espegren, Fredrik, 1989, et al. (författare)
  • Potential tellurium deposits in the BWR containment during a severe nuclear accident
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 146
  • Tidskriftsartikel (refereegranskat)abstract
    • The release of fission products to the environment is one of the concerns with nuclear power. During an accident, the most likely released are the volatile fission products i.e., tellurium. To evaluate the behavior of tellurium in the event of an accident, it was heated under different conditions (oxidizing, inert, reducing; both dry and humidified). The formed vapor was transported to surfaces (aluminum, copper, zinc) at room temperature that can be found in the BWR-containment. All formed deposits were examined for morphology and species. Moreover, the content of sodium hydroxide liquid traps following the metal surfaces and filter was also investigated. In these traps, the highest amount of tellurium was found under humid-reducing followed by humid-oxidizing conditions. In the deposit removed from the zinc surface acquired under the latter conditions, elemental analysis observed zinc, indicating a possible reaction between tellurium and zinc. The corresponding trap showed significant amounts of zinc. (C) 2020 The Author(s). Published by Elsevier Ltd.
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68.
  • Estévez-Albuja, S., et al. (författare)
  • Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 136
  • Tidskriftsartikel (refereegranskat)abstract
    • Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.
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69.
  • Fichot, F., et al. (författare)
  • Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 119, s. 36-45
  • Tidskriftsartikel (refereegranskat)abstract
    • The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).
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70.
  • Fokau, Andrei, et al. (författare)
  • A source efficient ADS for minor actinides burning
  • 2010
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:4, s. 540-545
  • Tidskriftsartikel (refereegranskat)abstract
    • Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.
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71.
  • Gajev, Ivan, et al. (författare)
  • Sensitivity analysis of input uncertain parameters on BWR stability using TRACE/PARCS
  • 2014
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 67, s. 49-58
  • Tidskriftsartikel (refereegranskat)abstract
    • The unstable behavior of Boiling Water Reactors (BWR), which is known to occur at certain power and flow conditions, could cause SCRAM and decrease the economic performance of the plant. For better prediction of BWR stability and understanding of influential parameters, two TRACE/PARCS models of Ringh-als-1 and Oskarshamn-2 BWRs were employed to perform a sensitivity study. Using the propagation of input errors uncertainty method's results, an attempt has been made to identify the most influential parameters affecting the stability. Furthermore, a methodology using the spearman rank correlation coefficient has been used to identify the most influential parameters on the stability parameters (decay ratio and frequency).
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72.
  • Gajev, Ivan, et al. (författare)
  • Space–time convergence analysis on BWR stability using TRACE/PARCS
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 51, s. 295-306
  • Tidskriftsartikel (refereegranskat)abstract
    • Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.
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73.
  • Gallego Marcos, Ignacio, et al. (författare)
  • Modelling of pool stratification and mixing induced by steam injectionthrough blowdown pipes
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 112, s. 624-639
  • Tidskriftsartikel (refereegranskat)abstract
    • Containment overpressure is prevented in a Boiling Water Reactor (BWR) by condensing steam into thepressure suppression pool. Steam condensation is a source of heat and momentum. Competition betweenthese sources results in thermal stratification or mixing of the pool. The interplay between the sources isdetermined by the condensation regime, steam mass flow rate and pool dimensions. Thermal stratificationis a safety issue since it limits the condensing capacity of the pool and leads to higher containmentpressures in comparison to a completely mixed pool with the same average temperature. The EffectiveHeat Source (EHS) and Effective Momentum Source (EMS) models were previously developed for predictingthe macroscopic effect of steam injection and direct contact condensation phenomena on the developmentof stratification and mixing in the pool. The models provide the effective heat and momentumsources, depending on the condensation regimes. In this work we present further development of theEHS/EMS models and their implementation in the GOTHIC code for the analysis of steam injection intocontainment drywell and venting into the wetwell through the blowdown pipes. Based on thePPOOLEX experiments performed in Lappeenranta University of Technology (LUT), correlations arederived to estimate the steam condensation regime and effective heat and momentum sources as functionsof the pool and steam injection conditions. The focus is on the low steam mass flux regimes withcomplete condensation inside the blowdown pipe or chugging. Validation of the developed methodswas carried out against the PPOOLEX MIX-04 and MIX-06 tests, which showed a very good agreementbetween experimental and simulation data on the pool temperature distribution and containmentpressure.
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74.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression Pool
  • Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
  • Tidskriftsartikel (refereegranskat)abstract
    • The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.
  •  
75.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Thermal stratification and mixing in a Nordic BWR pressure suppression pool
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 132, s. 442-450
  • Tidskriftsartikel (refereegranskat)abstract
    • The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.
  •  
76.
  • Galushin, Sergey, et al. (författare)
  • Sensitivity and Uncertainty Analysis of the Vessel Lower Head Failure Mode and Melt Release Conditions in Nordic BWR using MELCOR Code
  • 2020
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 135
  • Tidskriftsartikel (refereegranskat)abstract
    • Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Melt release mode into a deep water pool located in the lower drywell is a major source of uncertainty for success of such strategy. The main goal of this work is to identify the major contributors to the uncertainty in prediction of vessel failure mode, timing and melt release conditions in Nordic BWR using MELCOR severe accident analysis code. There is a forest of control rod guide tubes (CRGTs) and instrumentation guide tubes (IGTs) in the lower head of a BWR. Failure of the penetrations is considered in this work along with the gross creep rupture of the vessel lower head. Modelling of penetrations failure in MELCOR code is based on parametric models, allowing a user to select different assumptions, such as temperature threshold for penetrations failure, modelling of penetrations assembly and respective failure modes, mode of solid and liquid debris ejection from the vessel. In this work we perform sensitivity analysis to the MELCOR modelling options and sensitivity parameters in unmitigated station blackout scenario (SBO). Results of analysis suggest that (i) vessel breach due to penetration failure is predicted to occur before creep-rupture failure of the vessel lower head, (ii) the mode of debris ejection from the vessel has the dominant effect on the likelihood of creep-rupture failure of the vessel lower head and debris ejection rate from the vessel.
  •  
77.
  • Ghione, Alberto, 1989, et al. (författare)
  • Uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 104, s. 28-41
  • Tidskriftsartikel (refereegranskat)abstract
    • An uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor (JHR) is presented. The JHR is a new material testing reactor under construction at CEA on the Cadarache site, France. The thermal-hydraulic system code CATHARE is applied to investigate the response of the reactor system to the scenario. The uncertainty and sensitivity study was based on a statistical methodology for code uncertainty propagation, and the 'Uncertainty and Sensitivity' platform URANIE was used. Accordingly, the input uncertainties relevant to the transient, were identified, quantified, and propagated to the code output. The results show that the safety criteria are not exceeded and sufficiently large safety margins exist. In addition, the most influential input uncertainties on the safety parameters were found by making use of a sensitivity analysis.
  •  
78.
  • Gonzalez-Pintor, Sebastian, 1983, et al. (författare)
  • High Order Finite Element Method for the Lambda modes problem on hexagonal geometry
  • 2009
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:9, s. 1450-1462
  • Tidskriftsartikel (refereegranskat)abstract
    • A High Order Finite Element Method to approximate the Lambda modes problem for reactors with hexagonal geometry has been developed. This method is based on the expansion of the neutron flux in terms of the modified Dubiner's polynomials on a triangular mesh. This mesh is fixed and the accuracy of the method is improved increasing the degree of the polynomial expansions without the necessity of remeshing. The performance of method has been tested obtaining the dominant Lambda modes of different 2D reactor benchmark problems. © 2009 Elsevier Ltd. All rights reserved.
  •  
79.
  • Helgesson, Petter, 1986-, et al. (författare)
  • Treating model defects by fitting smoothly varying model parameters : Energy dependence in nuclear data evaluation
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 120, s. 35-47
  • Tidskriftsartikel (refereegranskat)abstract
    • The fitting of models to data is essential in nuclear data evaluation, as in many other fields of science. The models maybe necessary for interpolation or extrapolation, but they are seldom perfect; there are model defects present which can result in severe biases and underestimated uncertainties. This work presents and investigates the idea to treat this problem by letting the model parameters vary smoothly with an input parameter. To be specific, the model parameters for neutron cross sections are allowed to vary with neutron energy. The parameter variation is limited by Gaussian processes, but the method should not be confused with adding a Gaussian process to the model. The performance of the method is studied using a large number of synthetic data sets, such that it is possible to quantitatively study the distribution of results compared to the underlying truth. There are imperfections in the results, but the method is seen to readily outperform fits without the energy dependent parameters. In particular, the estimates of uncertainty and correlations are much better. Hence, the method seems to offer a promising route for future treatment of model defects, both for nuclear data and elsewhere.
  •  
80.
  • Hellesen, C., 1980-, et al. (författare)
  • Benchmark and demonstration of the CHD code for transient analysis of fast reactor systems
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 109, s. 712-719
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper the dynamic thermal hydraulic fast reactor simulation code CHD is presented. The code is built around a scriptable object-oriented framework in the programming language Python to be able to flexibly describe different reactor geometries including thermal-hydraulics models of an arbitrary number of coolant channels as well as pumps, heat-exchangers and pools etc. In addition, custom objects such as the Autonomous Reactivity Control (ARC) system for enhanced passive safety are modeled in detail. In this paper we compare the performance of the CHD code with other similar fast reactor dynamics codes using a benchmark study of the European Sodium cooled Fast Reactor (ESFR). The results agree well, both qualitatively and quantitatively with the code benchmark. In addition, we demonstrate the code's ability to simulate the long-term asymptotic behavior of a neutronically shut down reactor in an unprotected loss of flow scenario using a model of the Advanced Burner Reactor (ABR). (C) 2017 Elsevier Ltd. All rights reserved.
  •  
81.
  • Hellesen, Carl, 1980-, et al. (författare)
  • Nuclear Spent Fuel Parameter Determination using Multivariate Analysis of Fission Product Gamma Spectra
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 886-895
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, we investigate the application of multivariate data analysis methods to the analysis of gamma spectroscopy measurements of spent nuclear fuel (SNF). Using a simulated irradiation and cooling of nuclear fuel over a wide range of cooling times (CT), total burnup at discharge (BU) and initial enrichments (IE) we investigate the possibilities of using a multivariate data analysis of the gamma ray emission signatures from the fuel to determine these fuel parameters. This is accomplished by training a multivariate analysis method on simulated data and then applying the method to simulated, but perturbed, data.We find that for SNF with CT less than about 20 years, a single gamma spectrum from a high resolution gamma spectrometer, such as a high-purity germanium spectrometer, allows for the determination of the above mentioned fuel parameters.Further, using measured gamma spectra from real SNF from Swedish pressurized light water reactors we were able to confirm the operator declared fuel parameters. In this case, a multivariate analysis trained on simulated data and applied to real data was used.
  •  
82.
  • Herb, Joachim, et al. (författare)
  • Sensitivity analysis in core diagnostics
  • 2022
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 178
  • Tidskriftsartikel (refereegranskat)abstract
    • In the CORTEX project, methods to simulate neutron flux oscillations were enhanced and machine-learning based tools to determine the causes of measured neutron flux oscillations were developed, using the results of simulations as training and validation data. For a selected combination of those methods and tools, several sensitivity analyses were performed to assess their robustness and trustworthiness. The neutron flux oscillations were simulated using the tool CORE SIM+. It calculates the three-dimensional field of the neutron flux oscillations, which can be used to determine the response of neutron detectors at given locations. For the sensitivity analysis, the neutron flux oscillations were assumed to be caused by the vibration of one fuel element. It was investigated how selected input parameters like the core loading pattern, the burn up of the fuel elements, the neutronic core data, the geometry details of the vibrating fuel element, the chosen detectors, and other noise source parameters like the amplitude of the fuel element vibrations, affect the simulated neutron flux oscillations. A three dimensional fully convolutional neural network had been developed and trained during the CORTEX project to determine the cause and location of perturbations causing given measurements of in-core detectors in pressurized water reactors. The robustness of this network was tested by applying it to the simulated detector readings created during the sensitivity analysis.
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83.
  • Hernandéz Solís, Augusto, 1980, et al. (författare)
  • Uncertainty and sensitivity analyses applied to the DRAGONv4.05 code lattice calculations and based on JENDL-4 data
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 230-245
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, multi-group microscopic cross-section uncertainties are propagated through the DRAGON (Version 4.05) lattice code in order to perform uncertainty analysis on k(infinity) and 2-group homogenized macroscopic cross-sections. The test case corresponds to a 17 x 17 PWR fuel assembly segment without poison at full power conditions. A statistical methodology is employed for such purposes, where cross-sections of certain isotopes of various elements belonging to the 172 groups DRAGLIB library format, are considered as normal random variables. This library was based on JENDL-4 data, because JENDL-4 contains a large amount of isotopic covariance matrices among the different major nuclear data libraries. Thus, multi-group uncertainty was computed for the different isotopic reactions by means of ERRORRJ. The preferred sampling strategy for the current study corresponds to the quasi-random Latin Hypercube Sampling (LHS). This technique allows a much better coverage of the input uncertainties than simple random sampling (SRS) because it densely stratifies across the range of each input probability distribution. In order to prove this, the uncertain input space was re-sampled 10 times, and it is shown that the variability of the replicated mean of the different k(infinity) samples is much less for the LHS case, than for the SRS case. The uncertainty assessment of the output space should be based on the theory of non-parametric multivariate tolerance limits, due to the fact that k(infinity) and some of the macroscopic cross-sections are correlated. Therefore, for 10 replicated samples each containing 100 elements, the total output sample is composed by 1000 calculations. This sample size is more than enough to infer that the multivariate output population is covered 95% with a 95% of confidence. On the other hand, statistical sensitivity analysis was performed in order to know which microscopic cross-section has the greatest impact on k(infinity) predictions. It was found that the fission cross-section of Uranium 235 is the dominant input parameter for this particular case, because the computed JENDL-4 variances for such reaction are very high at thermal and resonant regions compared to other variances that for instance, can be computed based on other nuclear libraries such as ENDF/B-VII.1
  •  
84.
  • Holcombe, Scott, et al. (författare)
  • A Novel gamma emission tomography instrument for enhanced fuel characterization capabilities within the OECD Halden Reactor Project
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 837-845
  • Tidskriftsartikel (refereegranskat)abstract
    • Gamma emission tomography is a method based on gamma-ray spectroscopy and tomographic reconstruction techniques, which can be used for rod-wise characterization of nuclear fuel assemblies without dismantling the fuel. By performing a large number of measurements of the gamma-ray flux intensity around a fuel assembly using a well-collimated gamma-ray detector, the internal source distribution in the assembly may be reconstructed using tomographic algorithms. If a spectroscopic detection system is used, different gamma-ray emitting isotopes can be selected for analysis, enabling nondestructive fuel characterization with respect to a variety of fuel parameters. In this paper, we describe a novel gamma emission tomography instrument, which has been designed, constructed and tested at the Halden Boiling Water Reactor (HBWR). The device will be used to characterize fuel assemblies irradiated in the HBWR as part of ongoing nuclear fuel research conducted within the OECD Halden Reactor Project (HRP). As compared to single-rod gamma scanning, where the fuel is dismantled and the gamma radiation from each rod is measured separately, handling time associated with characterizing the fuel can be significantly reduced when using the gamma emission tomography device. Furthermore, because gamma emission tomography enables rod-wise fuel characterization without dismantling, even instrumented experimental fuel assemblies may be characterized repeatedly throughout the fuel's lifetime, with limited risk of damaging the fuel or its instrumentation. Accordingly, the capabilities of fuel characterization within the OECD HRP are expected to be strongly enhanced by the deployment of this device. Here, the gamma-tomographic method and the experimental setup are demonstrated through experimental measurements of the fuel stack and gas plenum regions of a nine-rod HBWR fuel assembly configuration, where four rods had a burnup of approximately 26 MWd/kgUO(2) and five rods had a burnup of approximately 50 MWd/kgUO(2). Tomographic images are presented, which show the applicability for assessment of fission gas contents in the gas plena and of fission products in the fuel stack. Furthermore, neutron activation products are analyzed, which give additional information on construction material properties.
  •  
85.
  • Holcombe, Scott, et al. (författare)
  • Feasibility of identifying leaking fuel rods using gamma tomography
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 334-340
  • Tidskriftsartikel (refereegranskat)abstract
    • In cases of fuel failure in irradiated nuclear fuel assemblies, causing leakage of fission gasses from a fuel rod, there is a need for reliable non-destructive measurement methods that can determine which rod is failed. Methods currently in use include visual inspection, eddy current, and ultrasonic testing, but additional alternatives have been under consideration, including tomographic gamma measurements.The simulations covered in this report show that tomographic measurements could be feasible. By measuring a characteristic gamma energy from fission gasses in the gas plenum, the rod-by-rod gamma source distribution within the fuel rod plena may be reconstructed into an image or data set which could then be compared to the predicted distribution of fission gasses, e.g. from the STAV code. Rods with significantly less fission gas in the plenum may then be identified as leakers.Results for rods with low fission gas release may, however, in some cases be inconclusive since these rods will already have a weak contribution to the measured gamma-ray intensities and for such rods there is a risk that a further decrease in fission gas content due to a leak may not be detectable. In order to evaluate this and similar experimental issues, measurement campaigns are planned using a tomographic measurement system at the Halden Boiling Water Reactor.
  •  
86.
  • Hoseyni, Seyed Mohsen, et al. (författare)
  • Melt infiltration through porous debris at temperatures above Solidification : Validation of analytical model
  • 2021
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 161, s. 108435-
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper investigates the dynamics of melt infiltration through a preheated porous debris bed which is of importance to severe accident modeling in nuclear power plants. Proper understanding of the flow physics and affecting parameters is needed to define flow regime(s) according to combination of the driving forces, i.e. capillary and gravity. A model development and validation therefore should consider various effects and competing mechanisms. After a careful study of the governing equations and scaling rules, a known analytical model is validated against existing experimental data from REMCOD experiment. The predictions of this model are in good agreement with the experimental data. Furthermore, a global sensitivity analysis identifies the most influential parameters and reveals the need for further experiments with different range of affecting parameters. The results underline the importance of permeability as the most influential parameter.
  •  
87.
  • Hou, Yandong, et al. (författare)
  • Effects of rolling motion on helical coil once-through steam generator thermal-hydraulic characteristics
  • 2023
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194, s. 110068-
  • Tidskriftsartikel (refereegranskat)abstract
    • Steam generators are essential for the safe, economical, and reliable operation of nuclear reactors. Helical Coil Once-through Steam Generators (HCOTSG) offer unique advantages over other types of steam generators for ship reactors or offshore platforms. To investigate the effect of rolling motion due to ocean conditions on the safe and economic operation of HCOTSG, a new model and a new code are developed in this paper. The program is developed and allows numerical simulation of HCOTSG under ocean conditions. The validity of the code was verified by comparing the simulation results with the design parameters of the Marine Reactor X (MRX) steam generator and the simulation results of other dedicated programs. The code was further used to perform the International Reactor Innovative and Secure (IRIS) transient operating conditions under typical rolling motions. Then, the influences of different swing directions, angles, periods, and positions of swing axes from the origin of the system operating parameters are analyzed. In the cases discussed in this paper, the following conclusions are obtained: (a) The direction of the swing significantly affects the system. The most dangerous situation is around the x-axis, and the case around the z-axis is the safest, in which the rolling situation has little effect on the system because the centripetal force is perpendicular to the tube wall, and the gravitational pressure drop is constant. (b) when the swing angle increases, except the fluctuation speed is faster, the fluctuation range of the parameters also increases. (c) when the swing period changes, the parameters’ fluctuation range also change (d) the orientation of the swing axis from the origin affects the magnitudes of the parameters’ changes (e) the distance of the swing axis from the origin affects the magnitude of the parameter variation and the farther the swing axis is from the origin, the greater the parameter fluctuation.
  •  
88.
  • Hou, Yandong, et al. (författare)
  • Numerical study on surface corrosion deposition of fuel elements and its influence on flow heat transfer
  • 2024
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 201
  • Tidskriftsartikel (refereegranskat)abstract
    • Corrosion of pressurized water reactors (PWR) in nuclear power plants can lead to serious safety hazards. This study aims to analyze the deposition of corrosion products using FLUENT software. Deposition models and thermal resistance models were developed, and the effects of deposits on the reactor's thermal–hydraulic characteristics were evaluated. Additionally, the impact of various parameters on deposition and thermal–hydraulic characteristics was examined. Results show that deposits accumulate extensively in the inlet section of the fuel cladding, while appearing as spot deposits in the outlet section. For deposit thicknesses below 30 μm, the surface temperature of the cladding gradually increases. However, when the thickness exceeds 30 μm, the surface temperature rapidly rises. Furthermore, the study reveals that the deposition amount decreases with increasing inlet flow velocity, exhibits an upward trend with higher inlet temperature, and increases with a higher wall heat flux density. This research provides important insights for understanding core deposition and thermal–hydraulic characteristics in nuclear reactor systems. It offers valuable guidance for enhancing safety and operational efficiency in nuclear power plants.
  •  
89.
  • Hou, Yandong, et al. (författare)
  • Thermal-hydraulic characteristics of helical coiled once-through steam generators in inclining condition of ocean
  • 2024
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 200
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper develops a one-dimensional thermal–hydraulic analysis program to simulate the main thermal–hydraulic parameter changes of helical coil once-through steam generator (HCOTSG) under inlet thermal–hydraulic parameters perturbation in inclining and vertical conditions. The inclining process is divided into the process of rotating from the vertical position to the inclining position and the stable inclining process. The rotating process is regarded as swaying motion. The swaying and inclining motions are achieved by modifying the momentum equation. What's more, the code was verified through experiments on two-phase heat transfer and friction pressure drop, as well as comparisons with the design parameters of the HCOTSG of the International Reactor Innovative and Secure (IRIS) reactor. The results of the simulation indicate that the direction of incline has an obvious impact on the safety of the HCOTSG, with inclination towards the y-axis having the greatest impact. In the stable inclining process, an increase in the angle of inclining results in a rise in primary-side outlet temperature and a reduction in heat transfer, while the secondary-side pressure drop increases. Furthermore, the HCOTSG underwent testing with eight types of transient perturbations, including four sudden perturbations and four linear perturbations. Results indicated that during sudden perturbations, apart from the secondary-side pressure drop, there were no substantial differences in simulated results between the inclined and vertical states at the same time during the transition process. However, during linear perturbations, due to the slow changes in parameters, the differences between the inclined and vertical states at the same time were more distinct. Regardless of the perturbation type, the inclined state led to a deterioration in the dynamic condition of the system compared to the vertical state.
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90.
  • Huang, Zi-Nan, et al. (författare)
  • Analysis of the stress field in the reactor vessel of the China Initiative Accelerator Driven System during postulated ULOF and UTOP transients
  • 2023
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194
  • Tidskriftsartikel (refereegranskat)abstract
    • The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The subcritical reactor in CiADS is a liquid Lead Bismuth Eutectic (LBE) cooled fast reactor. When the reactor core is in operation, the LBE coolant will directly contact and corrode the inner surface of reactor vessel. Due to the high temperature, the corrosion will be more severe. If the stress on the reactor vessel exceeds the limit, the plastic deformation will occur, leading to the generation and expansion of defects and cracks, and the safety of the reactor will be affected. Therefore, evaluating the stress field of the reactor vessel under different operating conditions is a very important research project. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS, and the ASME Code was used as stress assessment standards. We can preliminarily prove that the stress assessments of the vessel during the postulated Unprotected Loss of Flow (ULOF) accidents satisfy the requirements of ASME Code. The limit reactivity insertion to protect the vessel from plastic deformation is 0.58$ in the postulated Unprotected Transient over Power (UTOP) accidents based on our current results. Therefore, we can preliminarily conclude that the current material selection and structural design of the reactor vessel in CiADS could survive most of the postulated transient accidents considering the stress effect.
  •  
91.
  • Huang, Zheng, et al. (författare)
  • Numerical investigation on quench of an ex-vessel debris bed at prototypical scale
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 122, s. 47-61
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.
  •  
92.
  • Huang, Zheng, et al. (författare)
  • Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 128, s. 330-340
  • Tidskriftsartikel (refereegranskat)abstract
    • In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.
  •  
93.
  • Hursin, Mathieu, et al. (författare)
  • Modeling noise experiments performed at AKR-2 and CROCUS zero-power reactors
  • 2023
  • Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 194
  • Tidskriftsartikel (refereegranskat)abstract
    • CORTEX is a EU H2020 project (2017-2021) devoted to the analysis of ’reactor neutron noise’ in nuclear reactors, i.e. the small fluctuations occurring around the stationary state due to external or internal disturbances in the core. One important aspect of CORTEX is the development of neutron noise simulation codes capable of modeling the spatial variations of the noise distribution in a reactor. In this paper we illustrate the validation activities concerning the comparison of the simulation results obtained by several noise simulation codes with respect to experimental data produced at the zero-power reactors AKR-2 (operated at TUD, Germany) and CROCUS (operated at EPFL, Switzerland). Both research reactors are modeled in the time and frequency domains, using transport or diffusion theory. Overall, the noise simulators managed to capture the main features of the neutron noise behavior observed in the experimental campaigns carried out in CROCUS and AKR-2, even though computational biases exist close to the region where the noise-inducing mechanical vibration was located (the so-called ”noise source”). In some of the experiments, it was possible to observe the spatial variation of the relative neutron noise, even relatively far from the noise source. This was achieved through reduced uncertainties using long measurements, the installation of numerous, robust and efficient detectors at a variety of positions in the near vicinity or inside the core, as well as new post-processing methods. For the numerical simulation tools, modeling the spatial variations of the neutron noise behavior in zero-power research reactors is an extremely challenging problem, because of the small magnitude of the noise field; and because deviations from a point-kinetics behavior are most visible in portions of the core that are especially difficult to be precisely represented by simulation codes, such as experimental channels. Nonetheless the limitations of the simulation tools reported in the paper were not an issue for the CORTEX project, as most of the computational biases are found close to the noise source.
  •  
94.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Commercial thorium fuel manufacture and irradiation: Testing (Th,Pu)O-2 and (Th,U)O-2 in the "Seven-Thirty" program
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 79-86
  • Tidskriftsartikel (refereegranskat)abstract
    • Thorium based fuels are being tested in the Halden Research Reactor in Norway with the aim of producing the data necessary for licensing of these fuels in today's light water reactors. The fuel types currently under irradiation are thorium oxide fuel with plutonium as the fissile component, and uranium fuel with thorium as an additive for enhancement of thermo-mechanical and neutronic fuel properties. Fuel temperatures, rod pressures and dimensional changes are monitored on-line for quantification of thermo-mechanical behavior and fission gas release. Preliminary irradiation results show benefits in terms of lower fuel temperatures, mainly caused by improved thermal conductivity of the thorium fuels. In parallel with the irradiation, a manufacturing procedure for thorium-plutonium mixed oxide fuel is developed with the aim to manufacture industrially relevant high-quality fuel pellets for the next phase of the irradiation campaign.
  •  
95.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Irradiation testing of enhanced uranium oxide fuels
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 125, s. 99-106
  • Tidskriftsartikel (refereegranskat)abstract
    • Enhanced uranium oxide fuel types are being tested in the Halden Research Reactor in Norway with the aim is to assess the effect that these enhancements have on fuel performance. Fuel temperatures, rod pressures and dimensional changes are being monitored online and an extensive post-irradiation examination programme is planned. Preliminary data show that fuel centerline temperatures can be lowered by addition of ThO2 to the fuel matrix, or by incorporating Cr or SiO2-TiO2 as a network structure within the fuel. In parallel, two types of cladding coatings are tested in order to investigate their in-core properties. No abnormal behaviour has been noted during the first 100 days of irradiation.
  •  
96.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Thorium as an additive for improved neutronic properties in boiling water reactor fuel
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 113, s. 470-475
  • Tidskriftsartikel (refereegranskat)abstract
    • This article treats the replacement of burnable absorbers with a fertile absorber in boiling water reactor fuel. The target is to improve the fuel economy while meeting the same safety demands as the currently used conventional uranium oxide (UOX) fuel. A candidate fertile absorber is Th-232, and this work investigates the impact of replacing part of the U-238 in UOX fuel with Th-232. Computer simulations have been carried out and comparisons made for fuel assemblies with fertile and burnable absorbers, loaded in the boiling water reactor Oskarshamn 3, using the tools and methods that are normally employed for reload design and safety evaluation for this reactor. The results show that power balance and shutdown margins can be improved at the cost of higher enrichment needs. Alternatively, the fuel can be designed to just fulfil the relevant safety criteria, giving slightly lower uranium needs, which may compensate for the increased enrichment costs.
  •  
97.
  • Isotalo, A. E., et al. (författare)
  • Preventing xenon oscillations in Monte Carlo burnup calculations by enforcing equilibrium xenon distribution
  • 2013
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 60, s. 78-85
  • Tidskriftsartikel (refereegranskat)abstract
    • Existing Monte Carlo burnup codes suffer from instabilities caused by spatial xenon oscillations. These oscillations can be prevented by forcing equilibrium between the neutron flux and saturated xenon distribution. The equilibrium calculation can be integrated to Monte Carlo neutronics, which provides a simple and lightweight solution that can be used with any of the existing burnup calculation algorithms. The stabilizing effect of this approach, as well as its limitations are demonstrated using the reactor physics code Serpent.
  •  
98.
  • Jansson, Peter, 1971-, et al. (författare)
  • A new methodology for thermal analysis of geological disposal of spent nuclear fuel using integrated simulations of gamma heating and finite element modeling
  • 2022
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 172
  • Tidskriftsartikel (refereegranskat)abstract
    • A new methodology is illustrated, where the evolution of temperature in a geological disposal system for spent nuclear fuel is estimated by integrated calculations of a spatially distributed gamma heating source with conventional finite element thermal transport modeling. A case with one canister loaded with fuel assemblies with a cooling time of 30 years in a KBS-3 type repository illustrates the methodology. For this particular case, the effect of including distributed gamma heating rate in the modeling has a small impact on the temperature distribution compared to the conventional case of heat generated locally in the canister, resulting in a small decrease of the maximum temperature in the canister. A large proportion of gamma heating occurs inside the outer boundary of the copper canister for this case. Other potential consequences of radiation escaping the canister are discussed.
  •  
99.
  • Jareteg, Klas, 1986, et al. (författare)
  • Coupled fine-mesh neutronics and thermal-hydraulics - modeling and implementation for PWR fuel assemblies
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 84, s. 244-257
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper we present a fine-mesh solver aimed at resolving in a coupled manner and at the pin cell level the neutronic and thermal-hydraulic fields. Presently, the tool considers Pressurized Water Reactor (PWR) conditions. The methods and implementation strategy are such that the coupled neutronic and thermal-hydraulic problem is formulated in a fully three-dimensional (3D) and fine mesh manner, and for steady-state situations. The solver is built on finite volume discretization schemes, matrix solvers and capabilities for parallel computing that are availablein the open source C++ library foam-extend-3.0. The angular neutron flux is determined with a multigroup discrete ordinates method (SN ), solved by a sweeping algorithm. The thermal-hydraulics is based on Computational Fluid Dynamics (CFD) models for the moderator/coolant mass, momentum, and energy equations, together with the fuel pin energy equation. The multiphysics coupling is solved by making use of an iterative algorithm, and convergence is ensured for both the separate equations and the coupled scheme. Since all the equations are implemented in the same software, all fields can be directly accessed in such a manner that external transfer and external mapping are avoided. The parallelization relies on a domain decomposition which is shared between the neutronics and the thermal-hydraulics. The latter allows to exchange the coupled data locally on each CPU, thus minimizing the data transfer. The code is tested on a quarter of a 15 × 15 PWR fuel lattice. The results show that convergence is successfully reached, and correct physical behaviors of all fields can be achieved with a reasonable computational effort.
  •  
100.
  • Jareteg, Klas, 1986, et al. (författare)
  • Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal–hydraulic calculations
  • 2014
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 68, s. 247-256
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper investigates the feasibility of developing a fine mesh coupled neutronic/thermal–hydraulic solver within the same computing platform for selected fuel assemblies in nuclear cores. As a first step in this developmental work, a Pressurized Water Reactor at steady-state conditions was considered. The system being simulated has a finite axial size, but is infinite in the radial direction. The platform used for the modeling is based on the open source C++ library OpenFOAM. The thermal–hydraulics is solved using the built-in SIMPLE algorithm for the mass and momentum fields of the fluid, complemented by an equation for the temperature field applied simultaneously to all the regions (i.e. fluid and solid structures). For the neutronics, a two-group neutron diffusion-based solver was developed, with sets of macroscopic cross-sections generated by the Monte Carlo code SERPENT. The meshing of the system was created by the open source software SALOME. Successful convergence of the neutronic and thermal–hydraulic fields was achieved, thus bringing the solution of the coupled problem to an unprecedented level of details. Most importantly, the true interdependence of the different fields is automatically guaranteed at all scales. In addition, comparisons with a coarse-mesh radial averaging of the thermal–hydraulic variables show that a coarse-mesh fuel temperature identical for all fuel pins can lead to discrepancies of up to 0.5% in pin powers, and of several tens of pcm in multiplication factor.
  •  
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