Zobrazeno 1 - 10 of 15 pro vyhledávání: '"Carol Ahnert"'
1
PROPAGATION OF NUCLEAR DATA UNCERTAINTIES FOR PWR CORE ANALYSIS
Autor: C. Holgado, Oscar Cabellos, E. Castro, Carol Ahnert
Publikováno v: Nuclear Engineering And Technology, ISSN 1738-5733, 2014-06, Vol. 46, No. 3
Archivo Digital UPM
Universidad Politécnica de Madrid
Nuclear Engineering and Technology, Vol 46, Iss 3, Pp 299-312 (2014)
An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for 235,238 U, 239 Pu, and the therm
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_dedup___::4f5c9ce1df0efcc48e4fe397d4722411
https://doi.org/10.5516/net.01.2014.709
Zobrazit plný text záznamu
2
Neighborhood-corrected interface discontinuity factors for multi-group pin-by-pin diffusion calculations for LWR
Autor: Carol Ahnert, N. García-Herranz, D. Cuervo, J. J. Herrero
Publikováno v: Annals of Nuclear Energy, ISSN 0306-4549, 2012-08, Vol. 46
Archivo Digital UPM
instname
Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transpor
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_dedup___::b08d5217df71924d6f136500565cfb8e
https://doi.org/10.1016/j.anucene.2012.03.022
Zobrazit plný text záznamu
3
The Analytic Coarse-Mesh Finite Difference Method for Multigroup and Multidimensional Diffusion Calculations
Autor: Carol Ahnert, J.M. Aragones, N. García-Herranz
Publikováno v: Nuclear Science and Engineering. 157:1-15
In this work we develop and demonstrate the analytic coarse-mesh finite difference (ACMFD) method for multigroup - with any number of groups - and multidimensional diffusion calculations of eigenvalue and external source problems. The first step in t
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_________::77282b29bff01e14a03bedbffcc48338
https://doi.org/10.13182/nse07-a2709
Zobrazit plný text záznamu
4
Advanced multi-physics simulation for reactor safety in the framework of the NURESAFE project
Autor: Nikola Kolev, Angel Papukchiev, Victor Sanchez, Olivier Marchand, Nicolas Crouzet, Carol Ahnert, Bruno Chanaron, Soeren Kliem
Publikováno v: Annals of Nuclear Energy, ISSN 0306-4549, 2015-10, Vol. 84
Archivo Digital UPM
instname
Annals of Nuclear Energy
Annals of Nuclear Energy, Elsevier Masson, 2015, 84, pp.166-177. ⟨10.1016/j.anucene.2014.12.013⟩
Annals of Nuclear Energy, 2015, 84, pp.166-177. ⟨10.1016/j.anucene.2014.12.013⟩
Annals of Nuclear Energy 84(2015), 166-177
International audience; Since some years, there is a worldwide trend to move towards "higher-fidelity" simulation techniques in reactor analysis. One of the main objectives of the research in this area is to enhance the prediction capability of the c
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_dedup___::23f7626a65e5d48e0f189eac9b13fb0d
http://oa.upm.es/40443/
Zobrazit plný text záznamu
5
Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5
Autor: Vanessa Aragonés-Ahnert, Oscar Cabellos, Carol Ahnert, J.M. Aragones, N. García-Herranz
Publikováno v: Nuclear Technology. 146:29-40
The purpose of this paper is first to discuss the methods developed in our three-dimensional pressurized water reactor core dynamics code SIMTRAN and its coupling to the system code RELAP-5 for general transient and safety analysis. Then, we summariz
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_________::75529f56dbee82c2e114cc03eeea5695
https://doi.org/10.13182/nt04-a3484
Zobrazit plný text záznamu
6
Analytic Coarse-Mesh Finite-Difference Method Generalized for Heterogeneous Multidimensional Two-Group Diffusion Calculations
Autor: N. García-Herranz, Carol Ahnert, J.M. Aragones, Oscar Cabellos
Publikováno v: Nuclear Science and Engineering. 144:23-35
In order to take into account in a more effective and accurate way the intranodal heterogeneities in coarse-mesh finite-difference (CMFD) methods, a new equivalent parameter generation methodology has been developed and tested. This methodology accou
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_________::ac99fd674aa212ad448dbc11bd0d4624
https://doi.org/10.13182/nse03-a2340
Zobrazit plný text záznamu
7
Methods and Performance of the Three-Dimensional Pressurized Water Reactor Core Dynamics SIMTRAN On-Line Code
Autor: José M. Aragonés, Oscar Cabellos, Carol Ahnert
Publikováno v: Nuclear Science and Engineering. 124:111-124
New reactor physics and computation methods have been developed in the three-dimensional pressurized water reactor (PWR) core dynamics SIMTRAN code for on-line surveillance and prediction. The accuracy of the coupled neutronic thermal-hydraulic solut
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_________::75edd6c03d3f4416f354807d1314428f
https://doi.org/10.13182/nse96-a24228
Zobrazit plný text záznamu
8
The analytic nodal diffusion solver ANDES in multigroups for 3D rectangular geometry: Development and performance analysis
Autor: Carol Ahnert, N. García-Herranz, Juan-Andrés Lozano, José-María Aragonés
Publikováno v: Annals of Nuclear Energy, ISSN 0306-4549, 2008-12, Vol. 35, No. 12
Archivo Digital UPM
Universidad Politécnica de Madrid
In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy g
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_dedup___::153e2adbc18f38d6e5d50dabbacb21c2
http://oa.upm.es/2298/
Zobrazit plný text záznamu
10
Fuel Management and Core Design Code Systems for Pressurized Water Reactor Neutronic Calculations
Autor: Carol Ahnert, J.M. Aragones
Publikováno v: Nuclear Technology. 69:350-367
A package of connected code systems for the neutronic calculations relevant in fuel management and core design of pressurized water reactors (PWRs) has been developed and applied for validation to ...
Externí odkaz: https://explore.openaire.eu/search/publication?articleId=doi_________::b6c2923308612a103df8f7c5cf933486
https://doi.org/10.13182/nt85-a33617
Zobrazit plný text záznamu

Vyhledávací nástroje:

Upřesnit hledání

Omezení vyhledávání
Plný text Recenzováno Digitální knihovna AV ČR
Zdroje