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ORNL-4191 - the Molten Salt Energy Technologies Web Site

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r.<br />

~<br />

~<br />

~<br />

Fig. 4.8. Axiol Distribution of Thermal Neutron Flux in <strong>the</strong> MSRE with 233U Fuel Loading.<br />

THREE RODS WITHDRAWN -<br />

........ ..........<br />

<strong>ORNL</strong> -DWG 67--11904<br />

. .... ................<br />

I<br />

,- ThREE HODS !4ITI-!ORAWN<br />

i<br />

CONTRCL I I<br />

ELEMENT -7 i-<br />

........ I I<br />

L...l ............. I ........ 1 ...... 1 1 !<br />

30 75 20 15 40 5 0 5 10 15 70 75 30<br />

RADIAL DlSl ANCE (in )<br />

Fig. 4.9. Radial Distributions of Thermal Neutron Flux in <strong>the</strong> MSRE with 233U Fuel Loading.<br />

flux, averaged over <strong>the</strong> volume of <strong>the</strong> fuel circii-<br />

lating system and normalized to 4 blw. As dis-<br />

cussed in <strong>the</strong> preceding section, <strong>the</strong>se magnitudes<br />

are normalized to <strong>the</strong> minimum critical uranium<br />

concentration. Note also that <strong>the</strong>se values include Mw.-'.)<br />

<strong>the</strong> "flux dilution" effect of <strong>the</strong> time <strong>the</strong> fuel<br />

spends in <strong>the</strong> external pipirig and heat exchanger.<br />

(The corresponding value of <strong>the</strong> <strong>the</strong>rmal flux,<br />

averaged only over <strong>the</strong> graphite-moderated region<br />

of <strong>the</strong> core, is 3.60 x lo1' neutrons cm" sec..'

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