| 1 |
K.S. Moon, N.Z. Cho, J.M. Noh, and S.G. Hong, 'Acceleration of the Analytic Function Expansion Nodal Method by Two-Factor Two-Node Nonlinear Iteration,' Nucl. Sci. Eng., 132, 194 (1999)
|
| 2 |
D.S. Kim and N.Z. Cho, 'Acceleration of Three-Dimensional AFEN Nodal Codes via Coarse Group Rebalance and Direct Matrix Inverse,' Proc. Int. Conf. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, p.168, Madrid, Spain, September 27-30, 1999, and references therein
|
| 3 |
T. Ueki, T. Mori, and M. Nakagawa, 'Error Estimates and Their Biases in Monte Carlo Eigenvalue Calculations,' Nucl. Sci. Eng., 125, 1 (1997)
|
| 4 |
R.N. Blomquist and E.M. Gelbard, 'Alternative Implementations of the Monte Carlo Power Method,' Nucl. Sci. Eng., 141, 85 (2003)
|
| 5 |
R.J. Brissenden and A.R. Garlick, 'Biases in the Estimation of and Its Error by Monte Carlo Methods,' Ann. Nucl. Energy, 13, 63 (1986)
DOI
ScienceOn
|
| 6 |
G.I. Bell and S. Glasstone, Nuclear Reactor Theory, Van Nostrand Reinhold, 1970
|
| 7 |
J.J. Duderstadt and L.J. Hamilton, Nuclear Reactor Analysis, John Wiley & Sons, 1976
|
| 8 |
E.E. Lewis and W.F. Miller, Jr., Computational Methods of Neutron Transport, John Wiley & Sons, 1984
|
| 9 |
R.J.J. Stamm'ler and M.J. Abbate, Methods of Steady-State Reactor Physics in Nuclear Design, Academic Press, 1983
|
| 10 |
J. Gan, Y. Xu, and T.J. Downar, 'A Matrix-Free Newton Method for Coupled Neutronics Thermal-Hydraulics Reactor Analysis,' Int. Conf. Nuclear Mathematical and Computational Sciences (M&C 2003), Gatlinburg, USA, April 6-11, 2003, CD-ROM, American Nuclear Society (2003)
|
| 11 |
C.J. Jackson, D.G. Cacuci, and H.B. Finnemann, 'Dimensionally Adaptive Neutron Kinetics for Multidimensional Reactor Safety Transients ？ II : Dimensionally Adaptive Switching Algorithms,' Nucl . Sci. Eng., 131, 164 (1999)
|
| 12 |
D. Kastanya and P.J. Turinsky, 'Development of a Newton-Krylov Solver in FORMOSA-B,' Trans. Am. Nucl. Soc., 86, 226, (2002)
|
| 13 |
D. Kastanya and P.J. Turinsky, 'Implementation of a Newton-BiCGSTAB Solver to Treat the Strong Non-Linearity in the FORMOSA-B Boiling Water Reactor Core Simulator Code,'
|
| 14 |
Int. Conf. Nuclear Mathematical and Computational Sciences (M&C 2003), Gatlinburg, USA, April 6-11, 2003, CD-ROM, American Nuclear Society (2003)
|
| 15 |
B. Ivanov, K. Ivanov, P. Groudev, M. Pavlova and V. Hadjiev, 'VVER-1000 Coolant Transient Benchmark ？ PHASE I (V1000CT-1), Volume I : Final Specifications,' NEA/NSC/DOC, 6 (2002)
|
| 16 |
C.J. Jackson, D.G. Cacuci, and H.B. Finnemann, 'Dimensionally Adaptive Neutron Kinetics for Multidimensional Reactor Safety Transients ？ 1 : New Features of RELAP5/PANBOX,' Nucl. Sci. Eng., 131, 143 (1999)
|
| 17 |
P.J. Turinsky, et al., 'NESTLE: A Few-Group Neutron Diffusion Equation Solver,' EGG-NRE-11406, Electric Power Research Center, North Carolina State University (1994)
|
| 18 |
R. Boer and A. Knoll, 'Sensitivity Studies for Main Steam Line Break Exercises 2 and 3 with RELAP5/PANBOX,' Nucl. Technol., 142, 137 (2003)
|
| 19 |
H.B. Finnemann and A. Galati, 'NEACRP 3-DLWR Core Transient Benchmark, Final Specifications,' NEACRP-L-335 (Rev. 1), Nuclear Energy Agency Committee on Reactor Physics (Jan. 1992)
|
| 20 |
K.N. Ivanov, et al., 'PWR MSLB Benchmark. Volume 1 : Final Specifications,' NEA/NSC/DOC (99)8, Nuclear Energy Agency/Nuclear Science Committee (Apr. 1999)
|
| 21 |
'Argonne Code Center : Benchmark Problem Book,' ANL-7416, Suppl. 2, Argonne National Laboratory (1977)
|
| 22 |
K.N. Ivanov and N.K. Todorova, 'Using the OECD/NRC Pressurized Water Reactor Main Steam Line Break Benchmark to Study Current Numerical and Computational Issues of Coupled Calculations,' Nucl. Technol., 142, 95 (2003)
|
| 23 |
K.S. Smith, 'Nodal Method Storage Reduction by Nonlinear Iteration,' Trans. Am. Nucl. Soc., 44, 265 (1983)
|
| 24 |
N.Z. Cho and J.M. Noh, 'Hybrid of AFEN and PEN Methods for Multigroup Diffusion Nodal Calculation,' Trans. Am. Nucl. Soc., 73, 438 (1995)
|
| 25 |
H.C. Lee and C.H. Kim, 'Unified Formulation of Nodal Expansion Method and Analytic Nodal Method Solutions to Two-Group Diffusion Equations,' Nucl. Sci. Eng., 138, 192 (2001)
|
| 26 |
H.C. Lee and C.H. Kim, 'Unified Nodal Method Formulation for Analytic Function Expansion Nodal Method Solution to Two-Group Diffusion Equations in Rectangular Geometry,' Nucl. Sci. Eng., 140, 137 (2002)
|
| 27 |
T.M. Sutton and B.N. Aviles, 'Diffusion Theory Methods for Spatial Kinetics Calculations,' Prog. Nucle. Energy, 30,119 (1996)
DOI
ScienceOn
|
| 28 |
P.R. Engrand, G.I. Maldonado, R. Al-Chalabi, and P.J. Turinsky, 'Nonlinear Iteration Strategy for NEM : Refinement and Extension,' Trans. Am. Nucl. Soc., 65, 221, (1992)
|
| 29 |
H.G. Joo, et al., 'User Manual for the PARCS-v2.3 Beta Kinetics Core Simulator Module,' Purdue University, March 2002
|
| 30 |
N.Z. Cho, et al, 'Improved Features and Verification of an AFEN Method Code in Hexagonal-Z 3-D Geometry for Neutron Diffusion Calculation,' submitted to American Nuclear Society Annual Meeting, San Diego, CA, June 2005
|
| 31 |
J.M. Noh and N.Z. Cho, 'Intranodal Burnup Gradient Correction in Analytic Function Expansion Nodal (AFEN) Method,' Proceedings of the International Conference on Mathematics and Computations, Reactor Physics, and Environmental Analysis, Vol.2, pp.1435-1444, April 1995, Portland, U.S.A.
|
| 32 |
K. Ott and D.A. Meneley, 'Accuracy of the Quasistatic Treatment of Spatial Reactor Kinetics,' Nucl . Sci. Eng., 36, 402 (1969)
|
| 33 |
A. Pautz and A. Birkhofer, 'A Time-Dependent Neutron Transport Code Coupled with the Thermal-Hydraulics Code ATHLET,' Proc. Int. Conf. New Frontiers of Nuclear Technology: Reactor Physics, Safety and High- Performance Computing (PHYSOR 2002), Seoul, Korea, October 7-10, 2002, CD-ROM, American Nuclear Society (2002)
|
| 34 |
A.F. Henry, 'The Application of Reactor Kinetics to the Analysis of Experiments,' Nucl . Sci. Eng., 3, 52 (1958)
|
| 35 |
K. Ott, 'Quasistatic Treatment of Spatial Phenomena in Reactor Dynamics,' Nucl . Sci. Eng., 26, 563 (1966)
|
| 36 |
W.H. Reed and K.F. Hansen, 'Alternating Direction Methods for the Reactor Kinetics Equations,' Nucl . Sci. Eng., 41, 431 (1970)
|
| 37 |
W.M. Stacey, Jr., Space-Time Nuclear Reactor Kinetics, Academic Press, 1969
|
| 38 |
S.W. Woo, N.Z. Cho, and J.M. Noh, 'The Analytic Function Expansion Nodal Method Refined with Transverse Gradient Basis Functions and Interface Flux Moments,' Nucl. Sci. Eng., 139, 156 (2001)
|
| 39 |
Y.R. Park and N.Z. Cho, 'Coarse-Mesh Angular Dependent Rebalance Acceleration of the Discrete Ordinate Transport Calculations,' Nucl. Sci. Eng., 148, 355 (2004)
|
| 40 |
Y.R. Park and N.Z. Cho, 'The MOC Neutron Transport Calculations Accelerated by Coarse-Mesh Angular Dependent Rebalance,' Proc. Korean Nucl. Soc., Autumn Mtg., Yongpyong, October 2004
|
| 41 |
D.S. Kim and N.Z. Cho, 'Kinetics Calculation Under Space-Dependent Feedback in Analytic Function Expansion Nodal Method via Solution Decomposition and Galerkin Scheme,' Nucl. Sci. Eng., 140, 267 (2002)
|
| 42 |
D.S. Kim and N.Z. Cho, 'The Analytic Function Expansion Nodal (AFEN) Method with Half-Interface Averaged Fluxes in Mixed Geometry Nodes for Analysis of Pebble Bed Modular Reactor (PBMR) Cores,' Journal of Nuclear Science and Technology, 40, 291 (2003)
DOI
ScienceOn
|
| 43 |
H. Finnemann, F. Bennewitz and M. Wagner, 'Interface Current Techniques for Multidimensional Reactor Calculations,' Atomkernenergie, 30, 123 (1977)
|
| 44 |
S.G. Hong and N.Z. Cho, 'Convergence Analysis of the Angular-Dependent Rebalance Iteration Method in X-Y Geometry,' Trans. Am. Nucl. Soc., 80, 115 (1999)
|
| 45 |
Y.A. Chao and N. Tsoulfanidis, 'Conformal Mapping and Hexagonal Nodal Methods-I: Mathematical Foundation,' Nucl. Sci. Eng., 121, 202 (1995)
|
| 46 |
J.M. Noh and N.Z. Cho, 'A New Approach of Analytic Basis Function Expansion to Neutron Diffusion Calculation,' Nucl. Sci. Eng., 116, 165 (1994)
|
| 47 |
N.Z. Cho and J.M. Noh, 'Analytic Function Expansion Nodal Method for Hexagonal Geometry,' Nucl. Sci. Eng., 121, 245 (1995)
|
| 48 |
N.Z. Cho, Y.H. Kim, and K.W. Park, 'Extension of Analytic Function Expansion Nodal Method to Multigroup Problems in Hexagonal-Z Geometry,' Nucl. Sci. Eng., 126, 35 (1997)
|
| 49 |
K.S. Smith, 'An Analytic Nodal Method for Solving the Two-Group, Multidimensional, Static and Transient Neutron Diffusion Equations,' Thesis, Nuclear Engineering, Massachusetts Institute of Technology (1979)
|
| 50 |
K.S. Smith, 'Spatial Homogenization Methods for Light Water Reactor Analysis,' PhD Thesis, Massachusetts Institute of Technology (1980)
|
| 51 |
Y. Azmy, 'Impossibility of Unconditional Stability and Robustness of Diffusive Acceleration Schemes,' American Nuclear Society Radiation Protection and Shielding Division Topical Meeting, p. 480, Nashville, TN, U.S.A., April 19-23, 1998
|
| 52 |
N.Z. Cho, G.S. Lee, C.J. Park, 'Partial Current-Based CMFD Acceleration of the 2D/1D Fusion method for 3D Whole-Core Transport Calculations,' Trans. Am. Nucl. Soc., 88, 594, (2003)
|
| 53 |
S.G. Hong and N.Z. Cho, 'A Rebalance Approach to Nonlinear Iteration for Solving the Neutron Transport Equations,' Ann. Nucl. Energy, 24, 147 (1997)
DOI
ScienceOn
|
| 54 |
K. Koebke, 'A New approach to Homogenization and Group Condensation,' Proc. Specialists' Mtg. Homogenization Methods in Reactor Physics, Lugano, Switzerland, November 13-15, 1978, IAEA-TECDOC-232, p. 303, International Atomic Energy Agency (1978)
|
| 55 |
J.Y. Cho, et al., 'Cell-Based CMFD Formulation for Acceleration of Whole-Core Method of Characteristics Calculations,' Journal of the Korean Nuclear Society, 34, 250 (2002)
과학기술학회마을
|
| 56 |
H.G. Joo, et al., 'Dynamic Implementation of the Equivalence Theory in the Heterogeneous Whole Core Transport Calculation,' Proc. Int. Conf. New Frontiers of Nuclear Technology: Reactor Physics, Safety and High- Performance Computing (PHYSOR 2002), Seoul, Korea, October 7-10, 2002, CD-ROM, American Nuclear Society (2002)
|
| 57 |
N.Z. Cho, C.J. Park, 'A Comparison of Coarse Mesh Rebalance and Coarse Mesh Finite Difference Accelerations for the Neutron Transport Calculations,' Int. Conf. Nuclear Mathematical and Computational Sciences (M&C 2003), Gatlinburg, USA, April 6-11, 2003, CD-ROM, American Nuclear Society (2003)
|
| 58 |
K.S. Smith, J.D. Rhodes, III, 'CASMO Characteristics Method for Two-Dimensional PWR and BWR Core Calculation,' Trans. Am. Nucl. Soc., 83, 294, (2000)
|
| 59 |
K.S. Smith, J.D. Rhodes, III, 'Full-Core, 2-D, LWR Core Calculations with CASMO-4E,' Proc. Int. Conf. New Frontiers of Nuclear Technology: Reactor Physics, Safety and High- Performance Computing (PHYSOR 2002), Seoul, Korea, October 7-10, 2002, CD-ROM, American Nuclear Society (2002)
|
| 60 |
E.E. Lewis, et al., Expert Group on 3-D Radiation Transport Benchmarks ？ Benchmark Specification for Deterministic MOX Fuel Assembly Transport Calculations without Spatial Homoginisation (3-D Extension C5G7 MOX), NEA/NSC/DOC (2003)6, 2003
|
| 61 |
S. Kosaka and T. Takeda, 'Diffusion-Like 3-D Heterogeneous Core Calculation with 2-D Characteristics Transport Correction by Non-Linear Iteration Technique,' Int. Conf. Nuclear Mathematical and Computational Sciences (M&C 2003), Gatlinburg, USA, April 6-11, 2003, CD-ROM, American Nuclear Society (2003)
|
| 62 |
Y. Azmy, T. A. Wareing, and J. Morel, 'Effect of Material Heterogeneity on the Performance of DSA for Even-Parity SN Methods,' International Conference on Mathematics and Computation, Reactor Physics, and Environmental Analysis in Nuclear Applications, p.55, Madrid, Spain, September 27-30, 1999
|
| 63 |
J. S. Warsa, T. A. Wareing, J. E. Morel, J. M. McGhee, and R. B. Lehoucq, 'Krylov Subspace Iterations for the Calculation of k-Eigenvalues with Sn Transport Codes,' Int. Conf. Nuclear Mathematical and Computational Sciences (M&C 2003), Gatlinburg, USA, April 6-11, 2003, CD-ROM, American Nuclear Society (2003)
|
| 64 |
J.Y. Cho, et al., 'Three-Dimensional Heterogeneous Whole Core Transport Calculation Employing Planar MOC Solutions,' Trans. Am. Nucl. Soc., 87, 234 (2002)
|
| 65 |
G.S. Lee, N.Z. Cho, and S.G. Hong, 'Acceleration and Parallelization of the Method of Characteristics for Lattice and Whole-Core Heterogeneous Calculations,' Proc. PHYSOR 2000, Session II-C, Pittsburgh, USA, May 7-11, 2000
|
| 66 |
M.A. Smith, et al., 'Whole-Core Neutron Transport Calculations Without Fuel-Coolant Homogenization,' Proc. PHYSOR 2000, Session II-C, Pittsburgh, USA, May 7-11, 2000
|
| 67 |
N.Z. Cho, G.S. Lee, C.J. Park, 'Fusion of Method of Characteristics and Nodal Method for 3-D Whole-Core Transport Calculation,' Trans. Am. Nucl. Soc., 86, 322 (2002)
|
| 68 |
N.Z. Cho, et al., 'Whole-Core Heterogeneous Transport Calculations and Their Comparison with Diffusion Results,' Trans. Am. Nucl. Soc., 83, 292 (2000)
|
| 69 |
N.Z. Cho, et al., 'Refinement of the 2-D/1-D Fusion Method for 3-D Whole-Core Transport Calculation,' Trans. Am. Nucl. Soc., 87, 417 (2002)
|
| 70 |
S.G. Hong and N.Z. Cho, 'CRX: A Code for Rectangular and Hexagonal Lattices Based on the Method of Characteristics,' Ann. Nucl. Energy, 25, 547 (1998)
DOI
ScienceOn
|
| 71 |
S. Kosaka and E. Saji, 'The Characteristics Transport Calculation for a Multi-Assembly System using Neutron Path Linking Technique,' Proc. Int. Conf. Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, p.1890, Madrid, Spain, September 27-30, 1999
|
| 72 |
J.R. Askew, 'A Characteristics Formulation of the Neutron Transport Equation in Complicated Geometries,' AEEW-R-1108, U.K. Atomic Energy Authority (1972)
|
| 73 |
M.J. Halsall, 'CACTUS, A Characteristics Solutions to the Neutron Transport Equations in Complicated Geometries,' AEEW-R-1291, U.K. Atomic Energy Authority (1980)
|
| 74 |
A. Leonard and C.T. McDaniel, 'Optimal Polar Angles and Weights,' Trans. Am. Nucl. Soc., 73, 171 (1995)
|
| 75 |
W.L. Filippone, S. Woolf, and R.J. Lavigne, 'Particle Transport Calculations with the Method of Streaming Rays,' Nucl. Sci. Eng., 77, 119 (1981)
|
| 76 |
T.M. Sutton, 'NODEX: A High Order NEM-Based Multigroup Nodal Code,' Proc. Topical Meeting on Advances in Nuclear Engineering Computations and Radiation Shielding, 38:1-3:11. Santa Fe, New Mexico, April 9-13 (1989)
|
| 77 |
J.D. Densmore and E.W. Larsen, 'Variational Variance Reduction for Monte Carlo Eigenvalue and Eigenfunction Problems,' Nucl. Sci. Eng., 146, 121 (2004)
|
| 78 |
M.L. Adams and E.W. Larsen, 'Fast Iterative Methods for Discrete-Ordinates Particle Transport Calculations,' Progress in Nuclear Energy, 40, 3 (2002)
DOI
ScienceOn
|
| 79 |
E.L. Wachspress, Iterative Solution of Elliptic Systems, Prentice-Hall, 1966
|
| 80 |
G. Verdu, R. Miro, D. Ginestar, and V. Vidal, 'The Implicit Restarted Arnoldi Method, An Efficient Alternative to Solve the Neutron Diffusion Equation,' Ann. Nucl. Energy, 26, 579, (1999)
DOI
ScienceOn
|
| 81 |
A.F. Henry, Nuclear Reactor Analysis, The MIT Press, 1975
|
| 82 |
S. Nakamura, Computational Methods in Engineering and Science, John Wiley & Sons, 1997
|
| 83 |
E.M. Gelbard and A.G. Gu, 'Biases in Monte Carlo Eigenvalue Calculations,' Nucl. Sci. Eng., 117, 1 (1994)
|
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