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Uncertainty evaluation for $$^{244}$$Cm production in spent fuel of light water reactor by using burnup sensitivity analysis

燃焼感度解析手法を用いた軽水炉使用済み燃料中の$$^{244}$$Cm生成量の不確かさ評価

大泉 昭人; 横山 賢治; 石川 眞; 久語 輝彦

Oizumi, Akito; Yokoyama, Kenji; Ishikawa, Makoto; Kugo, Teruhiko

The uncertainty evaluation for the minor-actinide production is important to assure the reliability of the basic database of heat generation and radioactivity from reactor spent fuel. To identify the cross-section improvement priority for nuclide, reaction and energy range, the present paper describes the evaluation methodology for effective uncertainty reduction of target nuclide production by using the burnup sensitivity coefficients and the covariance of nuclear data. As a typical instance, the $$^{244}$$Cm production is focused on. The objects of uncertainty analysis are MOX and UO$$_{2}$$ of a pressurized water reactor, so that we can clarify the difference of the uncertainties between them. From the result, it is found that the nuclides near $$^{244}$$Cm on the burnup chain such as $$^{243}$$Am and $$^{242}$$Pu are important to produce $$^{244}$$Cm in both fuel types. In addition, it is confirmed the priority of $$^{243}$$Am, $$^{242}$$Pu and $$^{241}$$Pu is higher than $$^{235}$$U and $$^{239}$$Pu. Finally, the accuracy improvement of $$^{243}$$Am capture in the thermal and resonance regions should take a higher-priority than in the fast region.

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