Bias factor method using random sampling technique

References

• Shibata K, Iwamoto O, Nakagawa T, et al. JENDL-4.0: a new library for nuclear science and engineering. J Nucl Sci Technol. 2011;48:1–30.
   Web of Science ®Google Scholar
• JENDL-4.0u [Internet]. Japan: Japan Atomic Energy Agency; 2012 Sep 14 [updated 2015 Sep 14; cited 2015 Sep 22]. Available from: http://wwwndc.jaea.go.jp/jendl/j40/update/
   Google Scholar
• Chadwick MB, Herman M, Obložinský P, et al. ENDF/B-VII.1 nuclear data for science and technology: cross sections, covariances, fission product yields and decay data. Nucl Data Sheets. 2011;112:2887–2996.
   Google Scholar
• Yamamoto A, Endo T, Tabuchi M, et al. AEGIS: an advanced lattice physics code for light water reactor analyses. Nucl Eng Technol. 2010;42:500–519.
   Web of Science ®Google Scholar
• Tatsumi M, Yamamoto A. Advanced PWR core calculation based on multi-group nodal-transport method in three-dimensional pin-by-pin geometry. J Nucl Sci Technol. 2003;40:376–387.
   Web of Science ®Google Scholar
• Kato S, Endo T, Yamamoto A, et al. Random sampling-based cross-section adjustment technique for LWR core analysis. Paper presented at: 2013 International Congress on Advances in Nuclear Power Plants (ICAPP'13); 2013 Apr 14–18; Jeju Island, Korea.
   Google Scholar
• Watanabe T, Endo T, Yamamoto A, et al. Cross section adjustment method based on random sampling technique. J Nucl Sci Technol. 2014;51:590–599.
   Web of Science ®Google Scholar
• Hara A, Takeda T, Kikuchi Y. SAGEP: two-dimensional sensitivity analysis code based on generalized perturbation theory. Ibaraki (Japan): Japan Atomic Energy Agency; 1984 ( JAERI-M 84-027).
   Google Scholar
• Chiba G, Tsuji M, Narabayashi T. Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data. J Nucl Sci Technol. 2013;50:751–760.
   Web of Science ®Google Scholar
• Kamei T, Yoshida T. Error due to nuclear data uncertainties in the prediction of large liquid-metal fast breeder reactor core performance parameters. Nucl Sci Eng. 1983;84:83–97.
   Web of Science ®Google Scholar
• Sano T, Takeda T. Generalized bias factor method for accurate prediction of neutronic characteristics. J Nucl Sci Technol. 2006;43:1465–1470.
   Web of Science ®Google Scholar
• Kugo T, Mori T, Takeda T. Theoretical study on new bias factor methods to effectively use critical experiments for improvement of prediction accuracy of neutronic characteristics. J Nucl Sci Technol. 2007;44:1509–1517.
   Web of Science ®Google Scholar
• Kugo T, Andoh M, Kojima K, et al. Prediction accuracy improvement of neutronics characteristics of a breeding light water reactor core by extended bias factor methods with use of FCA-XXII-1 critical experiments. J Nucl Sci Technol. 2008;45:288–303.
   Web of Science ®Google Scholar
• Efron B. Bootstrap methods: another look at the Jackknife. Ann Stat. 1979;7:1–26.
   Web of Science ®Google Scholar
• Efron B, Tibshirani RJ. An introduction to the bootstrap. New York (NY): Chapman & Hall; 1993.
   Google Scholar
• Endo T, Watanabe T, Yamamoto A. Confidence interval estimation by bootstrap method for uncertainty quantification using random sampling method. J Nucl Sci Technol. 2015;52:993–999.
   Web of Science ®Google Scholar
• Muirhead RJ. Aspects of multivariate statistical theory. New York (NY): John Wiley & Sons; 1982.
   Google Scholar
• Yokoyama K, Ishikawa M, Kugo T. Extended cross-section adjustment method to improve the prediction accuracy of core parameters. J Nucl Sci Technol. 2012;49:1165–1174.
   Web of Science ®Google Scholar
• CASMO-4 A Fuel Assembly Burn-up Program. User's manual. Waltham (MA): Studsvik Scandpower Inc; 2009 (SSP-09/443-U Rev 0).
   Google Scholar
• SIMULATE-3 advanced three-dimensional two-group reactor analysis code. Waltham (MA): Studsvik Scandpower Inc; 2009 (SSP-09/447-U Rev 0).
   Google Scholar
• Endo T, Ohori K, Yamamoto A. Application of the robust design concept for fuel loading pattern. J Nucl Sci Technol. 2011;48:1077–1086.
   Web of Science ®Google Scholar
• Yamamoto A, Yasue Y, Endo T, et al. Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix. J Nucl Sci Technol. 2013;50:966–978.
   Web of Science ®Google Scholar