Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

[1]  David A. Brown,et al.  ENDF-6 Formats Manual: Data Formats and Procedures for the Evaluated Nuclear Data Files , 2018 .

[2]  Axel Hoefer,et al.  MOCABA: A GENERAL MONTE CARLO - BAYES PROCEDURE FOR IMPROVED PREDICTIONS OF INTEGRAL FUNCTIONS OF NUCLEAR DATA , 2014, 1411.3172.

[3]  Axel Hoefer,et al.  Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations , 2014, 1411.0834.

[4]  Carolina Ahnert Iglesias,et al.  Propagation of nuclear data uncertainties for PWR core analysis , 2014 .

[5]  William A. Wieselquist,et al.  PSI Methodologies for Nuclear Data Uncertainty Propagation with CASMO-5M and MCNPX: Results for OECD/NEA UAM Benchmark Phase I , 2013 .

[6]  Arjan J. Koning,et al.  Modern Nuclear Data Evaluation with the TALYS Code System , 2012 .

[7]  Bradley T Rearden,et al.  Development of a statistical sampling method for uncertainty analysis with SCALE , 2012 .

[8]  M. Klein,et al.  Interaction of loading pattern and nuclear data uncertainties in reactor core calculations , 2012 .

[9]  N. M. Larson,et al.  ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data , 2011 .

[10]  Andreas Pautz,et al.  Uncertainty Analyses with Nuclear Covariance Ddata in Reactor Core Ccalculations , 2011 .

[11]  Robert E. MacFarlane,et al.  Methods for Processing ENDF/B-VII with NJOY , 2010 .

[12]  M. Herman,et al.  ENDF-6 Formats Manual Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF/B-VII Written by the Members of the Cross Sections Evaluation Working Group Last Revision Edited by , 2009 .

[13]  Arjan J. Koning,et al.  Towards sustainable nuclear energy: Putting nuclear physics to work , 2008 .

[14]  R. Macfarlane,et al.  The NJOY Nuclear Data Processing System , 2008 .

[15]  H. Glaeser,et al.  Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation. IAEA Safety Report Series , 2008 .

[16]  B. Rearden,et al.  Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques , 2004 .

[17]  Philip F. Rose,et al.  ENDF-6 Formats Manual , 1997 .

[18]  R.E. MacFarlane,et al.  The NJOY Nu-clear Data Processing System Version 91 , 1994 .

[19]  J. F. Briesmeister MCNP-A General Monte Carlo N-Particle Transport Code , 1993 .

[20]  Igor Zmijarevic,et al.  APOLLO II: A User-Oriented, Portable, Modular Code for Multigroup Transport Assembly Calculations , 1988 .

[21]  A. Crespo,et al.  Validation of the Pressurized Water Reactor Core Analysis System SEANAP-86 with Measurements in Tests and Operation , 1988 .

[22]  J. M. Aragones,et al.  A Linear Discontinuous Finite Difference Formulation for Synthetic Coarse-Mesh Few-Group Diffusion Calculations , 1986 .

[23]  J. Aragonés,et al.  Fuel management and core design code systems for pressurized water reactor neutronic calculations , 1985 .

[24]  A. Gandini,et al.  A generalized perturbation method for bi-linear functionals of the real and adjoint neutron fluxes , 1967 .

[25]  L. N. Usachev Perturbation theory for the breeding ratio and for other number ratios pertaining to various reactor processes , 1964 .

[26]  M. Humi,et al.  MULTI-GROUP CONSTANTS FROM INTEGRAL DATA , 1964 .

[27]  M. Salvatores,et al.  ANALYSIS OF INTEGRAL DATA FOR FEW-GROUP PARAMETER EVALUATION OF FAST REACTORS , 1964 .