| 1. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Comparison of the U.S. NRC PARCS Core Neutronics Simulator Against In-Core Detector Measurements for LWR Applications
- 2012
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The safety analysis of a Nuclear Power Plant (NPP) is based on the application of complex computer codes that are able to simulate the physical behaviour of the system under normal operations and abnormal conditions. Therefore, such codes must be extensively and continuously verified and validated in order to demonstrate their reliability.In this context, the current report presents an assessment study for the U.S. NRC 3-D neutronic core simulator PARCS, and it includes an evaluation of the performances of the code for LWRs applications. For this purpose, the cores of the Swedish Ringhals-3 Pressurized Water Reactor (PWR) unit and the Forsmark-2 Boiling Water Reactor (BWR) unit were modeled with PARCS. As regards the cross-sections needed for this kind of calculations, they were prepared by following a special procedure developed by the present authors since core material data were only available in the format of library and restart files created by the SIMULATE-3 neutronic core simulator. Correspondingly, a new cross-section interface was developed and verified by the Division of Nuclear Engineering, Chalmers University of Technology, in order to convert the SIMULATE-3 data into data suitable for PARCS. Thereafter, the PARCS models developed for Ringhals-3 and Forsmark-2 were used for neutronic core analyses, at different operating conditions, along several fuel cycles. The results achieved from these simulations were then compared against the axial power and the radial power distribution estimated from the measurements that were provided by the owners of the plants.In the PWR case, the PARCS simulations predict satisfactorily both the core axial power profile and the core radial power distribution, although, in some cases, the deviations between calculated and measured data exhibit trends that need further investigations. For instance, the PARCS simulation at the beginning of a fuel cycle seems to overestimate the power in the center of the core and to underestimate the power at the periphery, whereas, at the end of a fuel cycle, the situation is opposite.In the BWR case, the core axial profile was predicted in a reasonable manner, but quite significant discrepancies for the radial power distribution was found. The current work suggests that such a disagreement might be due to the inability of PARCS to properly model multiple composition control rods. In fact the largest deviations in the computed power from the measurements were observed for those fuel assemblies placed in the neighborhood of control rods.
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| 2. |
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| 3. |
- Ghione, Alberto, 1989, et al.
(författare)
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On the prediction of single-phase forced convection heat transfer in narrow rectangular channels
- 2014
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Ingår i: Proceedings of the 10th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-10), Okinawa, Japan, December 14-18, 2014, on CD-ROM, Paper No. 1116.
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Konferensbidrag (refereegranskat)abstract
- In this paper, selected heat transfer correlations for single-phase forced convection are assessed for the case of narrow rectangular channels. The work is of interest in the thermal-hydraulic analysis of the Jules Horowitz Reactor (JHR), which is a research reactor under construction at CEA-Cadarache (France).In order to evaluate the validity of the correlations, about 300 tests from the SULTAN-JHR database were used. The SULTAN-JHR program was carried out at CEA-Grenoble and it includes different kinds of tests for two different vertical rectangular channels with height of 600 mm and gap of 1.51 and 2.16 mm. The experimental conditions range between 2 - 9 bar for the pressure; 0.5 - 18 m/s for the coolant velocity and 0.5 - 7.5 MW/m2 for the heat flux (whose axial distribution is uniform). Forty-two thermocouples and eight pressure taps were placed at several axial locations, measuring wall temperature and pressure respectively.The analysis focused on turbulent flow with Reynolds numbers between 5.5 x 103 - 2.4 x 105 and Prandtl numbers between 1.5 - 6. It was shown that standard correlations as the Dittus-Boelter and Seider-Tate significantly under-estimate the heat transfer coefficient, especially at high Reynolds number.Other correlations specifically designed for narrow rectangular channels were also taken into account and compared. The correlation of Popov-Petukhov in the form suggested by Siman-Tov still under-estimates the heat transfer coefficient, even if slight improvements could be seen. A better agreement for the tests with gap equal to 2.16 mm could be found with the correlation of Ma and the one of Liang. However the heat transfer coefficient when the gap is equal to 1.51 mm could not be predicted accurately. Furthermore these correlations were based on data at low Reynolds numbers (up to 13000) and low heat flux, so the use of them for SULTAN-JHR may be questionable.According to the authors’ knowledge, existing models of heat transfer coefficient in narrow channels have not been developed for high Reynolds number and high heat fluxes. Therefore, a new modified version of the Dittus-Boelter correlation was derived from a best-fitting of the SULTAN-JHR data with a multiple linear regression approach.The current study highlights that the channel geometry can impact the heat transfer. In particular, a reduction in gap size leads to an enhancement in the heat transfer coefficient.
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| 4. |
- Jareteg, Klas, 1986, et al.
(författare)
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Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal–hydraulic calculations
- 2014
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 68, s. 247-256
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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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| 5. |
- Jareteg, Klas, 1986, et al.
(författare)
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Influence of an SN solver in a fine-mesh neutronics/thermal- hydraulics framework
- 2014
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Ingår i: Proceedings of the International Conference on Physics of Reactors, PHYSOR 2014.
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Konferensbidrag (refereegranskat)abstract
- In this paper a study on the influence of a neutron discrete ordinates (SN) solver within a fine-mesh neutronic/thermal-hydraulic methodology is presented. The methodology consists of coupling a neutronic solver with a single-phase fluid solver, and it is aimed at computing the two fields on a three-dimensional (3D) sub-pin level. The cross-sections needed for the neutron transport equations are pre-generated using a Monte Carlo approach. The coupling is resolved in an iterative manner with full convergence of both fields. A conservative transfer of the full 3D information is achieved, allowing for a proper coupling between the neutronic and the thermal-hydraulic meshes on the finest calculated scales. The discrete ordinates solver is benchmarked against a Monte Carlo reference solution for a two-dimensional (2D) system.The results confirm the need of a high number of ordinates, giving a satisfactory accuracy in keff and scalar flux profile applying S16 for 16 energy groups. The coupled framework is used to compare the SN implementation and a solver based on the neutron diffusion approximation for a full 3D system of a quarter of a symmetric, 7x7 array in an infinite lattice setup. In this case, the impact of the discrete ordinates solver shows to be significant for the coupled system, as demonstrated in the calculations of the temperature distributions.
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| 6. |
- Jareteg, Klas, 1986, et al.
(författare)
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INVESTIGATION OF THE POSSIBILITY TO USE A FINE-MESH SOLVER FOR RESOLVING COUPLED NEUTRONICS AND THERMAL-HYDRAULICS
- 2013
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Ingår i: International Conference on Mathematics and Computational Methods Applied to Nuclear Science & Engineering (M&C 2013) Sun Valley, Idaho, USA, May 5-9, 2013, on CD-ROM. - 9781627486439 ; 3, s. 2002-2013
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Konferensbidrag (refereegranskat)abstract
- The development of a fine-mesh coupled neutronic/thermal-hydraulic solver is touched upon in this paper. The reported work investigates the feasibility of using finite volume techniques to discretize a set of conservation equations modeling neutron transport, fluid dynamics, and heat transfer within a single numerical tool. With the long-term objective of developing fine-mesh computing capabilities for a few selected fuel assemblies in a nuclear core, this preliminary study considers an infinite array of a single fuel assembly having a finite height. Thermal-hydraulic conditions close to the ones existing in PWRs are taken as a first test case. The neutronic modeling relies on the diffusion approximation in a multi-energy group formalism, with cross-sections pre-calculated and tabulated at the sub-pin level using a Monte Carlo technique. The thermal-hydraulics is based on the Navier-Stokes equations, complemented by an energy conservation equation. The non linear coupling terms between the different conservation equations are fully resolved using classical iteration techniques. Early tests demonstrate that the numerical tool provides an unprecedented level of details of the coupled solution estimated within the same numerical tool and thus avoiding any external data transfer, using fully consistent models between the neutronics and the thermal-hydraulics.
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| 7. |
- Nordlund, Anders, 1964, et al.
(författare)
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Building a comprehensive nuclear education in Sweden
- 2013
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Ingår i: Conference on Nuclear Training and Education 2013, CONTE 2013: An International Forum for Discussion of Issues Facing Nuclear Energy Training and Education. - 9781627480130 ; , s. 40-
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Konferensbidrag (refereegranskat)
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| 8. |
- Vinai, Paolo, 1975, et al.
(författare)
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Modelling of a self-sustained density wave oscillation and its neutronic response in a three-dimensional heterogeneous system
- 2014
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 67, s. 41-48
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Tidskriftsartikel (refereegranskat)abstract
- The main types of instabilities encountered in commercial Boiling Water Reactors (BWRs) are global and/or regional oscillations. In addition to those, pure Density Wave Oscillations (DWOs) have also been observed in some operating BWRs. These oscillations are particularly challenging from a modelling viewpoint because of the radially strongly localized character of the perturbation and of the corresponding neutronic response. In this paper, the features of a recently developed numerical tool, named CORE SIM, are taken advantage of. More specifically, this tool has the ability to estimate in the frequency domain the spatial and energy distribution of the stationary fluctuations of the neutron flux in any three-dimensional heterogeneous system. The perturbations should be directly defined in terms of fluctuations of the macroscopic cross-sections. In this study, the fluctuations in the macroscopic cross-sections are obtained by first modelling a boiling channel exhibiting a DWO with the US NRC RELAP5 code, and by thereafter converting the fluctuations of the coolant density along the channel into fluctuations of the macroscopic cross-sections using the Studsvik Scandpower CASMO-4E code. The RELAP5 and CASMO-4 models are representative of a typical BWR fuel assembly. The conditions modelled in RELAP5 were adjusted in order to obtain self-sustained DWOs. The axial distribution of the amplitude and phase of the fluctuations observed in the coolant density from the RELAP5 simulations are thus converted into fluctuations of the macroscopic cross-sections via CASMO-4E, and fed into a CORE SIM model representative of a heterogeneous BWR. The CORE SIM simulations in turn allow estimating the three-dimensional effects of a self-sustained DWO in a BWR core. More specifically, the axially dependent amplitude and phase of the variation of the coolant flow are properly accounted for, and the properties of the relative induced neutron fluctuations throughout the core are assessed.
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| 9. |
- Vinai, Paolo, 1975, et al.
(författare)
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Propagation of void fraction uncertainty measures in the RETRAN-3D simulation of the Peach Bottom turbine trip
- 2011
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 38:2-3, s. 358-370
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Tidskriftsartikel (refereegranskat)abstract
- The paper describes the propagation of void fraction uncertainty, as quantified by employing a novel methodology developed at Paul Scherrer Institut, in the RETRAN-3D simulation of the Peach Bottom turbine trip test. Since the transient considered is characterized by a strong coupling between thermal-hydraulics and neutronics. the accuracy in the void fraction model has a very important influence on the prediction of the power history and, in particular, of the maximum power reached. It has been shown that the objective measures used for the void fraction uncertainty, based on the direct comparison between experimental and predicted values extracted from a database of appropriate separate-effect tests, provides power uncertainty bands that are narrower and more realistic than those based, for example, on expert opinion. The applicability of such an approach to best estimate, nuclear power plant transient analysis has thus been demonstrated.
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