Change search
CiteExportLink to record
Permanent link

Direct link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf
Novel Diagnostics and Computational Methods of Neutron Fluxes in Boiling Water Reactors
Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
2010 (English)Doctoral thesis, comprehensive summary (Other academic)
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 In addition, two methods for determining the neutron flux in the control rod when it is withdrawn from the core are presented.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%.

Place, publisher, year, edition, pages
Uppsala: Acta Universitatis Upsaliensis , 2010. , p. 55
Series
Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology, ISSN 1651-6214 ; 786
Keywords [en]
Void, BWR, neutron fluxes, channel bow, MCNP, 2D/3D
National Category
Physical Sciences
Identifiers
URN: urn:nbn:se:uu:diva-133238ISBN: 978-91-554-7943-5 (print)OAI: oai:DiVA.org:uu-133238DiVA, id: diva2:360604
Public defence
2010-12-16, Häggsalen, Ångströmlaboratoriet, Lägerhyddsvägen 1, Uppsala, 10:15 (English)
Opponent
Supervisors
Note
Felaktigt tryckt som Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology 715Available from: 2010-11-23 Created: 2010-11-03 Last updated: 2011-03-21Bibliographically approved
List of papers
1.
The record could not be found. The reason may be that the record is no longer available or you may have typed in a wrong id in the address field.
2. A Method for Channel-Bow Indication by Neutron Flux Measurements
Open this publication in new window or tab >>A Method for Channel-Bow Indication by Neutron Flux Measurements
2010 (English)Conference paper, Published paper (Refereed)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:uu:diva-132857 (URN)
Conference
PHYSOR 2010 – Advances in Reactor Physics to Power the Nuclear Renaissance
Available from: 2010-10-27 Created: 2010-10-27 Last updated: 2011-01-13Bibliographically approved
3. Simulations and Models of Neutron Fluxes in BWRs Intended for Depletion Calculations of Withdrawn Control Rods
Open this publication in new window or tab >>Simulations and Models of Neutron Fluxes in BWRs Intended for Depletion Calculations of Withdrawn Control Rods
Show others...
2011 (English)In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 177, no 3, p. 221-229Article in journal (Refereed) Published
Abstract [en]

Models of the neutron flux shape in a withdrawn control rod in a boiling water reactor (BWR) bottom reflector have been constructed from simulations with the Monte Carlo code MCNP. These neutron flux models are intended for determining absorber depletion and fast fluence accumulation for withdrawn control rods with nodal codes. So-called G-factors are created for coupling the neutron flux models to a conventional nodal code via the core bottom neutron flux. The neutron flux models and G-factors are created for three different neutron energies, and their dependence on various parameters such as blanket enrichments, Hf and B(4)C control rod absorber, and depletion and reflector geometry is investigated. The neutron flux models and G-factors are found to be very insensitive; the neutron flux models predict the simulated neutron flux in the withdrawn control rod from MCNP over a variety of reflector configurations with an error < 3.0%. This implies that the neutron flux models constructed in this paper are generally applicable for BWR reflectors and control rods not fundamentally deviating from the designs investigated in this paper.

National Category
Physical Sciences
Identifiers
urn:nbn:se:uu:diva-132859 (URN)000288103600004 ()
Available from: 2010-10-27 Created: 2010-10-27 Last updated: 2017-12-12Bibliographically approved
4. Homogenization of Cross Sections and Computation of Discontinuity Factors for a Real 3D BWR Bottom Reflector for Comparison with Lattice and Nodal Codes
Open this publication in new window or tab >>Homogenization of Cross Sections and Computation of Discontinuity Factors for a Real 3D BWR Bottom Reflector for Comparison with Lattice and Nodal Codes
Show others...
2012 (English)In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 177, no 1, p. 1-7Article in journal (Refereed) 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.

National Category
Natural Sciences
Identifiers
urn:nbn:se:uu:diva-132861 (URN)000298446100001 ()
Available from: 2010-10-27 Created: 2010-10-27 Last updated: 2017-12-12Bibliographically approved
5. Investigation of Axial Power Gradients near an Unsymmetrical Control Blade Tip
Open this publication in new window or tab >>Investigation of Axial Power Gradients near an Unsymmetrical Control Blade Tip
Show others...
(English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed) Submitted
Identifiers
urn:nbn:se:uu:diva-132863 (URN)
Available from: 2010-10-27 Created: 2010-10-27 Last updated: 2017-12-12Bibliographically approved

Open Access in DiVA

fulltext(2706 kB)1243 downloads
File information
File name FULLTEXT01.pdfFile size 2706 kBChecksum SHA-512
5d447fcd76968324512960c565560befd045b6a8792f7042e77c3c152c90d5f3f3e44b4af22353f51c2df2b435432b92d3da002d7a397866adb02447281be13b
Type fulltextMimetype application/pdf
Buy this publication >>

By organisation
Applied Nuclear Physics
Physical Sciences

Search outside of DiVA

GoogleGoogle Scholar
Total: 1243 downloads
The number of downloads is the sum of all downloads of full texts. It may include eg previous versions that are now no longer available

isbn
urn-nbn

Altmetric score

isbn
urn-nbn
Total: 2086 hits
CiteExportLink to record
Permanent link

Direct link
Cite
Citation style
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf