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Technical Paper

Uncertainty Quantification of LWR Core Characteristics Using Random Sampling Method

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Pages 160-174 | Published online: 13 May 2017

REFERENCES

  • D. G. CACUCI, Sensitivity and Uncertainty Analysis: Theory, Vol. 1, Chapman and Hall/CRC, Boca Raton, Florida (2003).
  • H. S. ABDEL-KHALIK, “Adaptive Core Simulation,” PhD Dissertation, North Carolina State University (2004).
  • M. A. JESSEE, “Cross-Section Adjustment Techniques for BWR Adaptive Simulation,” PhD Dissertation, North Carolina State University (2008).
  • A. YAMAMOTO et al., “Uncertainty Estimation of Core Safety Parameters Using Cross-Correlations of Covariance Matrix,” J. Nucl. Sci. Technol, 50, 966 (2013); http://dx.doi.org/10.1080/00223131.2013.820155.
  • B. KRZYKACZ et al., “A Software System for Probabilistic Uncertainty and Sensitivity Analysis of Results from Computer Models,” Proc. Int. Conf. Probabilistic Safety Assessment and Management (PSAM-II), San Diego, California, March 1994.
  • M. KLEIN et al., “Influence of Nuclear Data Uncertainties on Reactor Core Calculations,” Kerntechnik, 76, 3, 174 (2011); http://dx.doi.org/10.3139/124.110148.
  • W. ZWERMANN et al., “Nuclear Data Uncertainty Analysis for a Fuel Assembly Criticality Benchmark,” Proc. Int. Conf. Nuclear Criticality, Edinburgh, United Kingdom, September 2011.
  • M. WILLIAMS et al., “A Statistical Sampling Method for Uncertainty Analysis with SCALE and XSUSA,” Nucl. Technol, 183, 515 (2013); http://dx.doi.org/10.13182/NT12-112.
  • W. WIESELQUIST et al., “Nuclear Data Uncertainty Propagation in a Lattice Physics Code Using Stochastic Sampling,” Proc. PHYSOR 2012, Knoxville, Tennessee, April 15–20, 2012, American Nuclear Society (2012).
  • I. PASICHNYK et al., “Effects of Nuclear Data Uncertainties on the NEA/OECD PWR MOX/UO2 Core Rod Ejection Benchmark,” Nucl. Technol, 183, 464 (2013); http://dx.doi.org/10.13182/NT183-464.
  • T. SAKAI et al., “Analysis of UAM Benchmark Problems Uncertainty Analysis of Core Problems Using Stochastic Sampling Method,” Proc. Fall Mtg. Atomic Energy Society of Japan, Hiroshima, Japan, September 19–21, 2012.
  • T. WATANABE et al., “Uncertainty and Correlation Estimation of Reload Safety Parameters ofPWR Using Random Sampling Method,” Trans. Am. Nucl. Soc, 109, 1365 (2013).
  • S. KATO et al., “Random Sampling-Based Cross-Section Adjustment Technique for LWR Core Analysis,” Proc. ICAPP 2013, Jeju, Korea, April 14–18, 2013.
  • M. D. McKAY, “Evaluating Prediction Uncertainty,” NUREG/CR-6311, LA-12915-MS, Los Alamos National Laboratory (1995).
  • T. TAKATA et al., “Uncertainty Correlation in Stochastic Safety Analysis of Natural Circulation Decay Heat Removal of Liquid Metal Reactor,” Proc. 13th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-13), Kanazawa, Japan, September 27–October 2, 2009.
  • K. KINOSHITA et al., “Uncertainty Quantification of Nuclear Characteristics Using Latin-Hypercube Sampling Method,” Proc. Int. Conf. PHYSOR 2014, Kyoto, Japan, September 28–October 3, 2014.
  • N. H. LARSEN, “Core Design and Operating Data for Cycle 1 and 2 of Peach Bottom 2,” NP-563, Electric Power Research Institute (1978).
  • N. H. LARSEN, “Core Design and Operating Data for Cycle 3 of Peach Bottom 2,” NP-971, Electric Power Research Institute (1981).
  • K. IVANOV et al., “Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of LWRs,” NEA/NSC/DOC(2007)23, Nuclear Energy Agency (2007).
  • R. E. MacFARLANE et al., “The NJOY Nuclear Data Processing System Version 91,” LA-12740-M, Los Alamos National Laboratory (Oct. 1994).
  • K. SHIBATA et al., “JENDL-4.0: A New Library for Nuclear Science and Engineering,” J. Nucl. Sci. Technol, 48,1 (2011); http://dx.doi.org/10.1080/18811248.2011.9711675.
  • “JENDL-4.0u”: http://wwwndc.jaea.go.jp/jendl/j40/update/, Japan Atomic Energy Agency (2012).
  • “CASMO-4: A Fuel Assembly Burn-Up Program. User’s Manual,” SSP-09/443-U Rev 0, Studvik Scandpower, Inc. (2009).
  • “SIMULATE-3 Advanced Three-Dimensional Two-Group Reactor Analysis Code,” SSP-09/447-U Rev 0, Studvik Scand-power, Inc. (2009).
  • T. ENDO et al., “Application of the Robust Design Concept for Fuel Loading Pattern,” J. Nucl. Sci. Technol, 48, 1077 (2011); http://dx.doi.org/10.1080/18811248.2011.9711792.

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