Zobrazeno 1 - 10
of 42
pro vyhledávání: '"CANFLEX"'
Publikováno v:
Annals of Nuclear Energy. 94:793-801
In this study, the use of the CANFLEX fuel bundle in the CANDU6 reactor has been investigated with high-fidelity Monte Carlo simulations in view of major safety parameters. The CANFLEX fuel bundle is designed such that it has a smaller linear power a
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::33809584a50c988b13604dbdb54b9f6b
https://doi.org/10.1016/j.anucene.2016.04.041
https://doi.org/10.1016/j.anucene.2016.04.041
Publikováno v:
Annals of Nuclear Energy. 94:313-324
This paper assesses a popular tube-based mechanistic critical heat flux model (Hewitt and Govan’s annular flow model (based on the model of Whalley et al.), and modifies and implements the model for bundle geometries. It describes the results of th
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::146093fc8396ac916866c3b9144173f6
https://doi.org/10.1016/j.anucene.2016.04.002
https://doi.org/10.1016/j.anucene.2016.04.002
Autor:
Jong Yeob Jeong, Jun Ho Bae
Publikováno v:
Nuclear Engineering and Technology, Vol 47, Iss 5, Pp 559-566 (2015)
The thermal–hydraulic characteristics for the CANadian Deuterium Uranium Flexible (CANFLEX)-burnable poison (BP) fuel channel, which is loaded with a BP at the center ring based on the CANFLEX-RU (recycled uranium) fuel channel, are evaluated and c
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_dedup___::6651bf6a63fb754624dd26df2be3d1d2
http://www.sciencedirect.com/science/article/pii/S1738573315001163
http://www.sciencedirect.com/science/article/pii/S1738573315001163
Publikováno v:
Nuclear Engineering and Design. 287:131-138
In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reacto
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::359a0db079d7ba037faaeb56613dba65
https://doi.org/10.1016/j.nucengdes.2015.03.010
https://doi.org/10.1016/j.nucengdes.2015.03.010
Publikováno v:
Proceedings of The 20th Pacific Basin Nuclear Conference ISBN: 9789811023163
In addition to electricity generation, CANDU6 reactors are used to produce Co-60, which is an important radioactive nuclide used in many industrial areas. In this study, an innovative way of producing Co-60 in CANDU6 reactors is introduced. Unlike th
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::0b080ba5ffd446a1431b9e0693607bf3
https://doi.org/10.1007/978-981-10-2317-0_64
https://doi.org/10.1007/978-981-10-2317-0_64
Publikováno v:
Nuclear Engineering and Design. 276:216-227
Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critic
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::249c1a7c9383f88dab3e9bc921cd8132
https://doi.org/10.1016/j.nucengdes.2014.06.004
https://doi.org/10.1016/j.nucengdes.2014.06.004
Publikováno v:
Nuclear Engineering and Design. 275:69-79
Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow dis
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::414b103e1695b3b00ecaaccd76c05899
https://doi.org/10.1016/j.nucengdes.2014.04.016
https://doi.org/10.1016/j.nucengdes.2014.04.016
Autor:
E.N. Onder, L.K.H. Leung
Publikováno v:
Nuclear Engineering and Design. 264:119-125
Boiling-Length-Average (BLA) Critical Heat Flux (CHF) values for the CANFLEX 1 bundle at cross-sectional average subcooled conditions have been evaluated using the ASSERT-PV subchannel code. The predicted BLA CHF values supplement experimental BLA CH
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::540530cbe11f407386ea87b6694fce47
https://doi.org/10.1016/j.nucengdes.2013.02.025
https://doi.org/10.1016/j.nucengdes.2013.02.025
Publikováno v:
Nuclear Engineering and Design. 240:1005-1012
A startup system and startup procedures are designed based on the subchannel analysis for CANDU-SCWR sliding pressure startup. Lookup tables are selected to predict the CHF and PDO heat transfer due to their wide application range. Plant parameters a
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::595c0dc0ef5ebaccc14d22678a539361
https://doi.org/10.1016/j.nucengdes.2009.12.037
https://doi.org/10.1016/j.nucengdes.2009.12.037
Publikováno v:
Progress in Nuclear Energy. 51:799-804
A cooperative study has been initiated at Xi'an Jiaotong University (XJTU) with Atomic Energy of Canada Limited (AECL) to develop a subchannel code ATHAS for preliminary analyses of flow and enthalpy distributions and cladding temperatures in CANDU f
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4587417200013191dc8bf1ccea54d7c0
https://doi.org/10.1016/j.pnucene.2009.05.004
https://doi.org/10.1016/j.pnucene.2009.05.004