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Homogenization of Cross Sections and Computation of Discontinuity Factors for a Real 3D BWR Bottom Reflector for Comparison with Lattice and Nodal Codes
Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
(Vattenfall Nuclear Fuel AB)
(Vattenfall AB)
Vise andre og tillknytning
2012 (engelsk)Inngår i: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 177, nr 1, s. 1-7Artikkel i tidsskrift (Fagfellevurdert) Published
Abstract [en]

Boiling water reactor (BWR) bottom reflector calculations in lattice codes such as CASMO are presently used only to produce accurate boundary conditions for core interfaces in nodal diffusion codes. Homogenized cross-section constants and discontinuity factors are calculated in one dimension (1-D) without the explicit presence of the control rod absorber. If the spatial flux in a BWR bottom reflector is required, for example, for depletion calculations of withdrawn control rods, the homogenization of the reflector must be based on a representation of the three-dimensional (3-D) geometry and material composition that is as true as possible. This paper investigates differences in cross-section and discontinuity factors from 1-D calculations in CASMO with 3-D Monte Carlo calculations of a realistic bottom reflector model in MCNP5. The cross-section and discontinuity factors from CASMO and MCNP5 are furthermore implemented in the nodal diffusion code SIMULATES to investigate the effect on the neutron fluxes in the bottom reflector. The results show that for the case investigated, the 1-D homogenization in CASMO5 produces a 26% overestimation of the homogenized thermal absorption cross section in the reflector and a 62% underestimation of the homogenized fast absorption cross section. These cross-section differences have essentially no impact on the neutron flux in the core but cause a 4.5% and 12.3% underestimation of the thermal and fast neutron flux, respectively, in the reflector region.

sted, utgiver, år, opplag, sider
2012. Vol. 177, nr 1, s. 1-7
HSV kategori
Identifikatorer
URN: urn:nbn:se:uu:diva-132861ISI: 000298446100001OAI: oai:DiVA.org:uu-132861DiVA, id: diva2:359345
Tilgjengelig fra: 2010-10-27 Laget: 2010-10-27 Sist oppdatert: 2017-12-12bibliografisk kontrollert
Inngår i avhandling
1. Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors
Åpne denne publikasjonen i ny fane eller vindu >>Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors
2010 (engelsk)Doktoravhandling, med artikler (Annet vitenskapelig)
Abstract [en]

The focus in this thesis is to improve knowledge of the BWR related uncertainties void, channel bow, and control rods. The presence of void determines the moderation of neutrons in BWRs. A high void fraction is less efficient in moderating neutrons than a low one. As a consequence, the ratio of thermal to fast neutrons is dependent on the surrounding void fraction. In this thesis, calculations with 2D/3D codes corroborate this dependence, the void correlation, to be linear and very robust to changes in different reactor parameters. The void fraction could be predicted from the ratio of simultaneously measured reaction rates from thermal and fast neutron detectors over the whole core with an uncertainty of ±1.5%. The only parameter found disturbing the void correlation significantly is channel bow. However, since channel bow is the only phenomenon found biasing the void correlation, it is found that the void prediction methodology can be used to indicate channel bow with a sensitivity of 4% per mm bow. Consequently, large channel bows could easily be detected. Increased knowledge of void fractions and channel bow could increase both safety and economy of nuclear power production.

This thesis also investigates how 2D/3D codes used in production perform in calculating detailed impact of control rods on pin powers and their ability to perform control rod depletion calculations in the reflector region. It is found that the axial resolution used in 3D nodal codes has very large impact on pin power gradients, i.e., using a standard nodal size of ~15 cm can cause underestimations of 50% in pin power gradients, which could lead to fuel damages. In addition, two methods for determining the neutron flux in the control rod when it is withdrawn from the core are presented. Both methods can be used in a 3D nodal code to reproduce the neutron flux in the reflector region with an uncertainty of ±3%.

sted, utgiver, år, opplag, sider
Uppsala: Acta Universitatis Upsaliensis, 2010. s. 55
Serie
Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology, ISSN 1651-6214 ; 786
Emneord
Void, BWR, neutron fluxes, channel bow, MCNP, 2D/3D
HSV kategori
Identifikatorer
urn:nbn:se:uu:diva-133238 (URN)978-91-554-7943-5 (ISBN)
Disputas
2010-12-16, Häggsalen, Ångströmlaboratoriet, Lägerhyddsvägen 1, Uppsala, 10:15 (engelsk)
Opponent
Veileder
Merknad
Felaktigt tryckt som Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology 715Tilgjengelig fra: 2010-11-23 Laget: 2010-11-03 Sist oppdatert: 2011-03-21bibliografisk kontrollert

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