| 1. |
- Insulander Björk, Klara L, 1982, et al.
(författare)
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Comparison of thorium-based fuels with different fissile components in existing boiling water reactors
- 2009
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Ingår i: Proc. Int. Congress on Advances in Nuclear Power Plants (ICAPP '09). ; 3, s. 1982-1987
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Konferensbidrag (refereegranskat)abstract
- Three different types of thorium based BWR fuel have been developed, in each of which thorium was combined with a different fissile component, the three components being reactor grade plutonium, uranium enriched to 20% in uranium 235 and pure uranium 233. A BWR nuclear bundle design, based on the geometrical fuel assembly design GE14, was developed for each of these fissile components. The properties and performance of the corresponding fuel assemblies were investigated via full core calculations carried out for an existing BWR and compared with the ones of an ordinary Low Enriched Uranium (LEU) fuel, which was developed for reference. The fuel assemblies and cores were designed to meet existing fuel design criteria, and were then analyzed with regards to reactivity coefficients, delayed neutron fractions, control rod worths and shutdown margins. The results show that all three alternatives seem to be feasible, although some difficulties remain with complying with the thermal limits, and with the moderator temperature and coolant void coefficients of the U-233 containing fuel being positive under some circumstances.
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| 2. |
- Ström, Henrik, 1981, et al.
(författare)
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Behaviour and stability of the two-fluid model for fine-scale simulations of bubbly flow in nuclear reactors
- 2015
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Ingår i: International Journal of Chemical Reactor Engineering. - : Walter de Gruyter GmbH. - 1542-6580 .- 2194-5748. ; 13:4, s. 449-459
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Tidskriftsartikel (refereegranskat)abstract
- In the present work, we formulate a simplistic two-fluid model for bubbly steam-water flow existing between fuel pins in nuclear fuel assemblies. Numerical simulations are performed in periodic 2D domains of varying sizes. The appearance of a non-uniform volume fraction field in the form of meso-scales is investigated and shown to be varying with the bubble loading and the domain size, as well as with the numerical algorithm employed. These findings highlight the difficulties involved in interpreting the occurrence of instabilities in two-fluid simulations of gas-liquid flows, where physical and unphysical instabilities are prone to be confounded. The results obtained in this work therefore contribute to a rigorous foundation in on-going efforts to derive a consistent meso-scale formulation of the traditional two-fluid model for multiphase flows in nuclear reactors.
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| 3. |
- Banati, Jozsef, 1960, et al.
(författare)
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Analysis of a loss of feedwater case at the Ringhals-3 NPP using RELAP5/PARCS coupled codes
- 2008
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Ingår i: 16th International Conference on Nuclear Engineering, ICONE16 2008; Orlando, FL; United States; 11 May 2008 through 15 May 2008. - 9780791848159 ; , s. 135-144
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005.At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions.Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.
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| 4. |
- Banati, Jozsef, 1960, et al.
(författare)
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Analysis of a loss of normal feedwater transient at the Ringhals-3 NPP using RELAP5/Mod.3.3
- 2010
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report gives an account on the development and validation of the RELAP5/Mod3.3 model of the Ringhals-3 pressurized water reactor against a Loss of Normal Feedwater Transient, which occurred on August 16, 2005. The 3rd unit of Ringhals Nuclear Power Plant comprises a 3-loops Westinghouse design pressurized water reactor on the Swedish West Coast.At first, the RELAP5 model is presented. All the 157 fuel assemblies are modeled individually in the code input. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops.The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, due to the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were challenged in this transient. The calculated values of the parameters show good agreement with the measured data.A parametric study was performed In order to evaluate the dependence of the steam generator level on the injected auxiliary feedwater flow. It indicated that the turbine driven auxiliary feedwater pump could possibly inject at a higher flowrate than its nominal value.The work was performed by the Department of Nuclear Engineering, Chalmers University of Technology in the framework of the Ringhals-3 power uprate project, supported by the Swedish Radiation Safety Authority (SSM). The ultimate goal of this project is to perform independent safety analyses of some limiting transients associated to the power uprate. The work carried out so far was targeted towards the development of state-of-the-art modelling capabilities for the Ringhals-3 unit.The present validational study is a Swedish contribution to the international Code Assessment and Maintenance Program (CAMP).
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| 5. |
- Banati, Jozsef, 1960, et al.
(författare)
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Development and validation of a coupled PARCS/RELAP5 model of the Ringhals-3 PWR
- 2007
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Ingår i: Proc. 15th Int. Conf. Nuclear Engineering (ICONE-15).
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the development of a coupled PARCS/RELAP5 model of the Swedish Ringhals-3 pressurized water reactor. The stand-alone PARCS and RELAP5 models are first presented. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. On the thermal-hydraulic side, each of the 157 fuel assemblies is modelled. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.Validation of the model against measured plant data at steady-state conditions is then summarized. Comparisons between calculated/measured parameters demonstrate that the model is able to correctly represent steady-state conditions of the plant.Finally, the validation of the model against measured transient plant data is described. The transient chosen for this validation task was a load rejection (“house-load”) transient, which occurred on January 8, 2005.
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| 6. |
- Banati, Jozsef, 1960, et al.
(författare)
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Validational exercises with RELAP5/PARCS coupled codes using Ringhals-3 plant data
- 2008
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Ingår i: Proc. Int. Conf. Physics of Reactors "Nuclear Power: A Sustainable Resource" (PHYSOR'08). - 9783952140956
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Konferensbidrag (refereegranskat)abstract
- The present study is part of a series of validation exercises, in which features of a coupled RELAP5/PARCS model is being assessed against various steady-state and transient plant data obtained at the Swedish Ringhals-3 NPP. The current work is focusing on the simulation of two specific off-normal situations, such as a Load Rejection and a Loss of Normal Feedwater case occurred in 2005. At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modelled individually in both codes. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes.The load rejection transient was initiated by disconnection of the electric grid, which resulted in a reduced power generation. The other analysed transient started with a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions.Capabilities of the RELAP5 code were more challenged in these transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analyses. The calculated values of the parameters show good agreement with the measured data.
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| 7. |
- Chambon, Amalia, 1986, et al.
(författare)
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VALIDATION OF HORUS3D/N AGAINST TRIPOLI-4®D FOR CORE DEPLETION CALCULATION OF THE JULES HOROWITZ REACTOR
- 2016
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Ingår i: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016; Paper No. 15947. - 9781510825734 ; 1, s. 140-151
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Konferensbidrag (refereegranskat)abstract
- The international Jules Horowitz material testing Reactor (JHR) is under construction at CEA Cadarache research center, in southern France. Its first criticality is foreseen by the end of the decade. In order to perform JHR design and safety studies, a specific neutronics calculation tool, HORUS3D/N, based on the deterministic codes APOLLO2 and CRONOS2 and on the European nuclear data library JEFF3.1.1, was developed to calculate JHR neutronics parameters taking into account fuel depletion: reactivity, power distribution, control rod reactivity worth, etc. Up to now, the biases and uncertainties on the different neutronics parameters computed with HORUS3D/N were assessed, in particular, by comparing HORUS3D/N deterministic calculations with reference route calculations based on APOLLO2-MOC and TRIPOLI-4®. The use for JHR of the recent Monte-Carlo TRIPOLI-4® in its new Depletion mode (TRIPOLI-4®D) will also allow providing biases for the main neutronics parameters under fuel depletion conditions. These biases will give a quantitative estimation of the impact of the approximations of the flux calculation in the deterministic route. This paper presents a contribution to the validation of HORUS3D/N based on the first comparisons between the calculations performed with APOLLO2-MOC and CRONOS2, and the ones from TRIPOLI-4®D. The study is performed on 2-D calculations for two different clusters in an infinite lattice configuration. It focuses on the main parameters of interest: isotopic concentrations, plate power distributions, reactivity, as functions of burnup. The results obtained show reasonable discrepancies with APOLLO2 calculation and allow to be confident on the APOLLO2.8/REL2005/CEA2005 package recommendations developed by CEA for light water reactor studies used in HORUS-3D/N. In particular, the main fuel isotopes are well predicted with TRIPOLI-4®D with discrepancies values lower than -1.5%.
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| 8. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Calculation of the eigenfunctions and corresponding eigenvalues of the two-group diffusion equation in heterogeneous systems
- 2007
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Ingår i: Proc. Joint Int. Topl. Mtg. Mathematics & Computation and Supercomputing in Nuclear Applications (M & C + SNA 2007). - 0894480596
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Konferensbidrag (refereegranskat)abstract
- This paper deals with the extension of an existing code developed at the Department of Nuclear Engineering, Chalmers University of Technology, to the calculation of the higher order eigenfunctions and corresponding eigenvalues of the two-group diffusion equations. More specifically, this code, originally meant for calculating the two-group static neutron flux and its adjoint for any two-dimensional heterogeneous system, is modified in order to calculate any higher mode of both the neutron flux and its adjoint. A one-dimensional two-region system is first considered for benchmark purposes, where the reference solution can be determined semi-analytically. The numerical solution is computed using a modified power iteration method. The agreement between the numerical solution and the reference solution is found to be excellent, both regarding the eigenfunctions and the eigenvalues, and both for the forward and adjoint problems. The modified power iteration method is then applied to the calculation of the different modes of the neutron flux and its adjoint in a two-dimensional heterogeneous representation of the Swedish Forsmark-1 BWR core corresponding to the 1996/1997 channel instability event.
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| 9. |
- Demaziere, Christophe, 1973
(författare)
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Description of the models and algorithms used in the CORE SIM neutronic tool
- 2011
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The development of an innovative neutronic tool is reported hereafter. The novelty of the tool resides in its versatility, since many different systems can be investigated and different kinds of calculations can be performed. More precisely, both critical systems and subcritical systems with an external neutron source can be studied, and static and dynamic cases in the frequency domain (i.e. for stationary fluctuations) can be considered. In addition, the tool has the ability to determine the different eigenfunctions of any nuclear core. For each situation, the static neutron flux, the different eigenmodes and eigenvalues, the first-order neutron noise, and their adjoint functions are estimated, as well as the effective multiplication factor of the system. The main advantages of the tool, which is entirely MatLab based, lie with the robustness of the implemented numerical algorithms, its high portability between different computer platforms and operative systems, and finally its ease of use since no input deck writing is required. The present version of the tool, which is based on two-group diffusion theory, is mostly suited to investigate thermal systems. Although the tool cannot be compared in terms of accuracy to existing core simulators, the definition of both the static and dynamic core configurations directly from the static macroscopic cross-sections and their fluctuations, respectively, makes the tool particularly well suited for research and education. This report describes the neutronic models and numerical algorithms implemented in the tool, whereas the validation and demonstration of the tool is reported in a companion report (Demazière, 2011a). The tool, for which a complete user’s manual exists (Demazière, 2011b), is freely available on direct request to the author of the present report.
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| 10. |
- Demaziere, Christophe, 1973
(författare)
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Development of computational methods and their applications for the analysis of nuclear power plants
- 2009
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Ingår i: International Journal of Nuclear Energy Science and Technology. - 1741-637X .- 1741-6361. ; 4:4, s. 287-298
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Tidskriftsartikel (refereegranskat)abstract
- A specificity of nuclear reactors is their multiphysics and multiscale character. The multiphysics nature comes from the interdependency between different fields governing the physics of such systems (neutron transport, heat transfer and fluid dynamics). The coupling between physical phenomena across various characteristic lengths, varying from the microscale to the macroscale, requires a multiscale treatment. Specific modelling techniques at the system level are thus required for the simulation of nuclear reactors and are presented in this paper. The use of such techniques for both time-independent and time-dependent simulations are dealt with, and examples of such simulations are presented. Time-independent simulations are mostly carried out for in-core fuel management purposes, i.e., for designing a core loading that allows running the reactor in a safe and economical manner. For the time-dependent simulations, two classes of problems are encountered. If the system undergoes small stationary fluctuations whereas the mean values of the variables remain constant, linear theory can be used to find the governing equations of such fluctuations. Due to the stationary character of the fluctuations, such calculations are more easily performed in the frequency domain. If the system undergoes large fluctuations and/or if the mean values of the variables are changing with time, the equations are to be solved in the time domain.
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