Zobrazeno 1 - 8
of 8
pro vyhledávání: '"Anitta Hämäläinen"'
Publikováno v:
Hovi, V, Taivassalo, V, Hämäläinen, A, Räty, H & Syrjälahti, E 2017, ' Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO ', Kerntechnik, vol. 82, no. 4, pp. 426-435 . https://doi.org/10.3139/124.110820
The 7th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system co
Publikováno v:
Arkoma, A, Hänninen, M, Rantamäki, K, Kurki, J & Hämäläinen, A 2015, ' Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant ', Nuclear Engineering and Design, vol. 285, pp. 1-14 . https://doi.org/10.1016/j.nucengdes.2014.12.023
In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been develo
Publikováno v:
Kliem, S, Danilin, S, Hämäläinen, A, Hádek, J, Keresztúri, A & Siltanen, P 2007, ' Qualification of coupled 3-D neutron-kinetic/thermal-hydraulic code systems by the calculation of main-steam-line-break benchmarks in an NPP with VVER-440 reactor ', Nuclear Science and Engineering, vol. 157, no. 3, pp. 280-298 . https://doi.org/10.13182/NSE07-A2728
Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydraulic system codes. These coupled codes can be used for the analysis of the whole reactor system. In the framework of the international association Ato
Autor:
Anitta Hämäläinen, Elina Syrjälahti
Publikováno v:
Syrjälahti, E & Hämäläinen, A 2006, ' HEXTRAN-SMABRE calculation of the VVER-1000 coolant transient-1 benchmark ', Progress in Nuclear Energy, vol. 48, no. 8, pp. 849-864 . https://doi.org/10.1016/j.pnucene.2006.06.007
The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps a
Publikováno v:
Daavittila, A, Hämäläinen, A & Kyrki-Rajamäki, R 2003, ' Effects of secondary circuit modeling on results of pressurized water reactor main steam line break benchmark calculations with new coupled code TRAB-3D/SMABRE ', Nuclear Technology, vol. 142, no. 2, pp. 116-123 . < http://www.ans.org/pubs/journals/nt/a_3377 >
All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finlan
Autor:
G. Hegyi, Frank-Peter Weiss, Sören Kliem, Anitta Hämäläinen, Riitta Kyrki-Rajamäki, S. Danilin, S. Langenbuch, S. Mittag, J. Hadek
Publikováno v:
Hämäläinen, A, Kyrki-Rajamäki, R, Mittag, R, Kliem, S, Weiss, F, Langenbuch, S, Danilin, S, Hadek, J & Hegyi, G 2002, ' Validation of coupled neutron kinetic/thermal-hydraulic codes : Part 2: Analysis of a VVER-440 transient (Loviisa-1) ', Annals of Nuclear Energy, vol. 29, no. 3, pp. 255-269 . https://doi.org/10.1016/S0306-4549(01)00039-1
Several three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal-hydraulic system codes. Under the auspices of the European Union's Phare programme these codes have been validated
Autor:
S. Danilin, S. Langenbuch, S. Nikonov, Anitta Hämäläinen, Sören Kliem, G. Hegyi, B. Krzykacz-Hausmann, J. Hadek, V. Khalimanchuk, A. Keresztúri, K.-D. Schmidt, A. Kuchin
Publikováno v:
Langenbuch, S, Krzykacz-Hausmann, B, Schmidt, K-D, Hegyi, G, Keresztúri, A, Kliem, S, Hadek, J, Danilin, S, Nikonov, S, Kuchin, A, Khalimanchuk, V & Hämäläinen, A 2005, ' Comprehensive uncertainty and sensitivity analysis for coupled code calculations of VVER plant transients ', Nuclear Engineering and Design, vol. 235, no. 2-4, pp. 521-540 . https://doi.org/10.1016/j.nucengdes.2004.09.003
The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics codes, is an important step to perform best-estimate calculations for plant transients of nuclear power plants. For applications in safety analysis, th
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::3938a9f51557e31ce3d25e0a71b325f7
https://cris.vtt.fi/en/publications/7a910f6e-9460-4b49-af01-148d1f8ab170
https://cris.vtt.fi/en/publications/7a910f6e-9460-4b49-af01-148d1f8ab170
Publikováno v:
Miettinen, J, Hämäläinen, A & Pekkarinen, E 2002, Generalized thermohydraulics module GENFLO for combining with the PWR core melting model, BWR recriticality neutronics model and fuel performance model . in 10th International Conference on Nuclear Engineering : Thermal Hydraulics . vol. 3, American Society of Mechanical Engineers (ASME), New York, pp. 665-678, 10th International Conference on Nuclear Engineering, ICONE 10, Arlington, Virginia, United States, 14/04/02 . https://doi.org/10.1115/ICONE10-22421
Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dyn
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::dda218bde6b8dd5a23e20653249b6917
https://cris.vtt.fi/en/publications/c2b95a0f-b730-4e7f-ab4b-c7d6080b5e4e
https://cris.vtt.fi/en/publications/c2b95a0f-b730-4e7f-ab4b-c7d6080b5e4e