JAEA-Conf 2011-002 - 日本原子力研究開発機構
JAEA-Conf 2011-002 - 日本原子力研究開発機構
JAEA-Conf 2011-002 - 日本原子力研究開発機構
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<strong>JAEA</strong>-<strong>Conf</strong> <strong>2011</strong>-<strong>002</strong><br />
<br />
<br />
Takuya Ito, Shogo Matsuoka, Shota Okui<br />
Nuclear Fuel Industries, Ltd. Tokai-works,<br />
Tokai-mura, Naka-gun, Ibaraki, 319-1196, Japan<br />
Email:t-itoh@nfi.co.jp<br />
To understand the characteristics of evaluated nuclear data from a view point of isotope<br />
concentration prediction, burnup calculations for high burnup BWR MOX fuels were performed based<br />
on several nuclear data libraries, including JENDL-4.0. The calculation results were compared to<br />
measurements and each other calculations.<br />
<br />
JENDL/AC-2008 1) was released in 2008 in Japan. In USA, ENDF/B-VII.0 2) was released in<br />
2006 and preparation of ENDF/B-VII.1 is progressing. In Europe, JEFF-3.1.1 3) was released in 2009.<br />
JENDL-4.0 4) was released in May 2010.<br />
In this study, isotope concentration prediction calculations for high burnup BWR MOX fuels<br />
with the new several evaluated nuclear data files are performed and compared to reference values.<br />
The result of C/E comparison to the results of radio-chemical analysis of the MALIBU 5,6) high<br />
burn up BWR MOX samples is shown for JENDL-3.2 7) , JENDL/AC-2008, ENDF/B-VII.0, JEFF-3.1 by<br />
using MVP-BURN 8) . MALIBU program is an international PIE program for high burnup fuel which was<br />
irradiated in commercial reactors.<br />
As for JENDL-4.0, C/C comparison of isotope concentration against the latest evaluated<br />
nuclear data files at high burn up BWR MOX condition is shown.<br />
Continuous energy Monte-Carlo code "MVP-BURN" is employed for burnup calculation to<br />
reduce both of geometry approximation and effective cross section approximation. In calculation, fuel<br />
assembly geometry is simulated exactly and irradiation data which are provided by MALIBU program<br />
are traced.<br />
<br />
MALIBU program 5,6) is an international PIE program for high burnup MOX and UOX fuels. In<br />
this program, various nuclides, 17 heavy metal nuclides and 34 fission products nuclides are subject of<br />
analysis. The sample matrix is shown in Table 1.