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Validation of a new CBH model in the nuclear core simulator POLCA7
2004 (English)Independent thesis Advanced level (professional degree), 20 credits / 30 HE creditsStudent thesis
Abstract [en]

When maintaining a nuclear reactor it is essential to keep track of the various reactions that take place there. The two most important processes are the power distribution and the neutron flux density distribution. Almost any alteration in the core results in changing the structure of the neutron flux. The flux, in turn, will eventually affect the power output and the burnup pattern for the fuel assemblies. At all times it is necessary to have accurate, up-to-the-minute information in order to run the nuclear core in an efficient and safe way. However, since continuous measurements for all reactions in all points are impossible simulations have to be performed. The control blade history effect (CBH effect) describes the redistribution of power inside a fuel assembly when a control rod is inserted during depletion. The CBH effect causes a decrease in the power output when the control rod is inserted, and gives rise to a reactivity peak immediately after the control rod is withdrawn. Eventually this peak fades out. Also, the CBH effect will lead to a highly heterogeneous power distribution over the fuel segments near the control rod region. Due to the control blade, or more commonly control rod, which is highly absorbant to thermal neutrons, major changes in the neutron spectrum take place. The thermal neutrons are the predominant source of the fission reaction in the nuclear core and thus changes to the neutron spectrum will greatly affect the power output. Therefore, a CBH model was implemented in the core simulator POLCA7 to correct for the CBH effect. The old model was separated from the rest of the POLCA7 models and should thus be regarded as an ad-hoc solution. The new, implicit, CBH model is partly implemented into the cross-section model in POLCA7, and corrects for deviations in average cross-sections as well as for side/edge cross-sections. This is done regardless of the presence of a control rod. In this master thesis we evaluate a new version of the CBH model, with respect to pin power, in version of the nodal code POLCA7 compared to the cross-section generator CASMO-4 as a reference program. We tested in a BWR- environment the FANP Atrium10B fuel assemblies with 7 burnable absorber pins in a 2x2 assembly configuration of different ages. One major challenge was to establish the 2x2 assembly configuration in both programs in order to make an accurate comparison. The new CBH model was tested for three cases, varying: 1) the depletion length with the control rod inserted, 2) the fuel assembly configurations and 3) the depletion void value. The test results in all evaluation cases were unambiguous: the new CBH model had larger pin power deviations compared to the old model. The errors were larger on pin level, control assembly level as well as on the whole fuel assembly configuration level. For the base configuration at control rod assembly level, with depletion length of 8 MWd/kgU and 40% void, the RMS error was 7,91% for the new CBH model compared to 1,24% for the old. Finally, the conclusion of this validation is that the new CBH model is not yet sufficient enough to compensate for the tilted power distribution over the fuel segments caused by the CBH-effect. A revision of the new terms implemented in the cross-section model regarding the new, implicit, CBH model is advised.

Place, publisher, year, edition, pages
Keyword [en]
Technology, nuclear core simulator, CBH, validation
Keyword [sv]
URN: urn:nbn:se:ltu:diva-57805ISRN: LTU-EX--04/274--SELocal ID: e6dc9c5f-8443-4d71-8773-f638d878e4beOAI: diva2:1031193
Subject / course
Student thesis, at least 30 credits
Educational program
Engineering Physics, master's level
Validerat; 20101217 (root)Available from: 2016-10-04 Created: 2016-10-04Bibliographically approved

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