Zobrazeno 1 - 10
of 22
pro vyhledávání: '"Rami Pohja"'
Publikováno v:
Pohja, R, Holmström, S, Auerkari, P & Nurmela, A 2017, ' Predicted life of P91 steel for cyclic high temperature service ', Materials at High Temperatures, vol. 34, no. 5-6, pp. 301-310 . https://doi.org/10.1080/09603409.2017.1383710
The P91 steel is widely used in high temperature components of power plants, and it is a candidate material for Gen-IV reactors. The P91 steel has relatively attractive mechanical and physical properties combined with resistance to stress corrosion c
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::bc716db35b897b8b5600a9ced8c463ce
https://doi.org/10.1080/09603409.2017.1383710
https://doi.org/10.1080/09603409.2017.1383710
Publikováno v:
Holmström, S, De Haan, F, Föhrer, U, Pohja, R & Janousek, J 2017, ' Engineering models for softening and relaxation of Gr. 91 steel in creep-fatigue conditions ', International Journal of Structural Integrity, vol. 8, no. 6, pp. 670-682 . https://doi.org/10.1108/IJSI-02-2017-0010
Purpose There are a number of different approaches for calculating creep-fatigue (CF) damage for design, such as the French nuclear code RCC-MRx, the American ASME III NH and the British R5 assessment code. To acquire estimates for the CF damage, tha
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::01ffd487d4c69318340bf22efe1e33cb
https://doi.org/10.1108/IJSI-02-2017-0010
https://doi.org/10.1108/IJSI-02-2017-0010
Autor:
Janne Heikinheimo, Caitlin Huotilainen (Hurley), Rami Pohja, Mykola Ivanchenko, Henri Loukusa
Publikováno v:
Heikinheimo, J, Huotilainen, C, Pohja, R, Ivanchenko, M & Loukusa, H 2019, Advanced cladding materials for accident tolerant fuels . in SYP2019 Proceedings . Finnish Nuclear Society, Nuclear Science and Technology Symposium, NST 2019, Helsinki, Finland, 20/10/19 . < https://ats-fns.fi/images/files/2019/syp2019/presentations/TSFM1_HLoukusa_AdvancedCladdingMaterialsForAccidentTolerantFuels_Tier1paper.pdf >
VTT Technical Research Centre of Finland-PURE
VTT Technical Research Centre of Finland-PURE
In terms of safety, nuclear fuel cladding is one of the main barriers in preventing the release of radioactive materials to the environment.The integrity of the fuel cladding is essential both in operation and in the long-term storage of spent nuclea
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=dedup_wf_001::f2658f5c1e17b4ccd3a67441fc178fe5
https://cris.vtt.fi/en/publications/ff4317df-d873-4dd9-a1e2-da917fa8077e
https://cris.vtt.fi/en/publications/ff4317df-d873-4dd9-a1e2-da917fa8077e
Autor:
Asko Arkoma, Timo Ikonen, Joonas Kättö, Henri Loukusa, Emmi Myllykylä, Rami Pohja, Ville Tulkki, Ville Valtavirta
Publikováno v:
Arkoma, A, Ikonen, T, Kättö, J, Loukusa, H, Myllykylä, E, Pohja, R, Tulkki, V & Valtavirta, V 2017, Physics and chemistry of nuclear fuel (PANCHO) . in J Hämäläinen & V Suolanen (eds), SAFIR2018-The Finnish Research Programme on Nuclear Power Plant Safety 2015-2018 : Interim Report . VTT Technical Research Centre of Finland, VTT Technology, vol. 294, pp. 192-202 .
VTT Technical Research Centre of Finland-PURE
VTT Technical Research Centre of Finland-PURE
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=dedup_wf_001::15e54ac3888e830da4cf80d03a90151e
https://cris.vtt.fi/en/publications/c31a9c4d-a5ed-4693-8d89-d93ec4558d69
https://cris.vtt.fi/en/publications/c31a9c4d-a5ed-4693-8d89-d93ec4558d69
Publikováno v:
Pohja, R, Nurmela, A, Moilanen, P & Holmström, S 2013, ' Multifunctional high precision pneumatic loading system (HIPS) for creep-fatigue testing ', Procedia Engineering, vol. 55, pp. 573-577 . https://doi.org/10.1016/j.proeng.2013.03.297
The first successful application of the high precision pneumatic loading system for advanced material testing was developed at VTT in 1995. The pneumatic servo-controlled loading system is capable of operating in a range of extreme conditions such as
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::2b68bda5e9d0eccb883c52d665357dff
https://doi.org/10.1016/j.proeng.2013.03.297
https://doi.org/10.1016/j.proeng.2013.03.297
Publikováno v:
Holmström, S, Pohja, R, Nurmela, A, Moilanen, P & Auerkari, P 2013, ' Creep and creep-fatigue behaviour of 316 stainless steel ', Procedia Engineering, vol. 55, pp. 160-164 . https://doi.org/10.1016/j.proeng.2013.03.236
The austenitic stainless steel 316 is of current interest as structural material for the future Gen IV nuclear power plants operating at high temperatures. Although 316 steel grades have been studied for the service conditions of current nuclear and
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::047243b8da8a09f0b60eef139832746e
https://doi.org/10.1016/j.proeng.2013.03.236
https://doi.org/10.1016/j.proeng.2013.03.236
Publikováno v:
Auerkari, P, Yli-Olli, S, Penttilä, S, Tuurna, S & Pohja, R 2016, ' Performance of high-temperature materials for efficient power plants : The waterside challenge ', Journal of Nuclear Engineering and Radiation Science, vol. 2, no. 3, 031009 . https://doi.org/10.1115/1.4032783
Supercritical (SC) service at high operating values aims for good plant efficiency, but the waterside oxidation resistance can then become life-limiting. In this paper, selected materials and modeling options are compared for life assessment under wa
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::1042cca2dae98b4bd389e1cb836f5b16
https://cris.vtt.fi/en/publications/65bbac90-8815-464c-a8ec-734196ad7ac3
https://cris.vtt.fi/en/publications/65bbac90-8815-464c-a8ec-734196ad7ac3
Autor:
Ville Tulkki, Ulla Ehrnstén, Pekka Moilanen, Juhani Rantala, Rami Pohja, Stefan Holmström, Santtu Huotilainen
Publikováno v:
Pohja, R, Moilanen, P, Rantala, J, Huotilainen, S, Tulkki, V, Ehrnstén, U & Holmström, S 2016, ' Multiaxial creep testing device for nuclear fuel claddings ', Athens Journal of Sciences, vol. 3, no. 3, pp. 227-240 . < http://www.athensjournals.gr/ajs/past/v3i3 >
VTT Technical Research Centre of Finland-PURE
VTT Technical Research Centre of Finland-PURE
Mechanical properties and creep resistance are among the key issues in the selection of fuel cladding materials for future Gen-IV nuclear reactor applications. As for today's nuclear reactors, the Fukushima accident has increased the interest in acci
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::d8064fae09d3afd6fd151057445db72e
https://cris.vtt.fi/en/publications/63c37e65-ed13-4090-90b9-5a12a68d79ef
https://cris.vtt.fi/en/publications/63c37e65-ed13-4090-90b9-5a12a68d79ef
Publikováno v:
Tuurna, S, Yli-Olli, S, Pohja, R & Penttilä, S 2016, Supercritical water oxidation and creep behaviour of boiler tube materials . in Proceedings of 8th International Conference on Advances in Materials Technology for Fossil Power Plants . ASM International, pp. 924-930, 8th International Conference on Advances in Materials Technology for Fossil Power Plants, Algarve, Portugal, 10/10/16 .
Scopus-Elsevier
Scopus-Elsevier
High efficiency in power generation is not only desirable because of economical reasons but also for enhanced environmental performance meaning reduced quantity of forming ash and emissions. In modern medium to large size plants, improvements require
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=dedup_wf_001::26be1123826b0f92c14a966b719f82d3
https://cris.vtt.fi/en/publications/6ad60dd3-087b-42d5-b36f-7f0fc87c76ac
https://cris.vtt.fi/en/publications/6ad60dd3-087b-42d5-b36f-7f0fc87c76ac
Autor:
Stefan Holmström, Rami Pohja
Publikováno v:
Pohja, R & Holmström, S B 2014, A comparison of creep-fatique assessment and modelling methods . in Proceedings of the 22th International Conference on Nuclear Engineering, ICONE22 : Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues . vol. 1, ICONE22-30640, American Society of Mechanical Engineers (ASME), 22nd International Conference on Nuclear Engineering, ICONE 22, Prague, Czech Republic, 7/07/14 . https://doi.org/10.1115/ICONE22-30640
Design codes, such as RCC-MRx and ASME III NH, for generation IV nuclear reactors use interaction diagram based method for creep-fatigue assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the all
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::36289eb447a367ab9b4dcc28790d8470
https://cris.vtt.fi/en/publications/14c0ae28-21f7-4faf-bdda-95525c4fccb1
https://cris.vtt.fi/en/publications/14c0ae28-21f7-4faf-bdda-95525c4fccb1