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As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming.

After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows:

(1) Derivation of the multigroup transport equations and the multigroup diffusion equations, with representative solution methods thereof.

(2) Elements of modern (now almost three decades old) diffusion nodal methods.

(3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues.

(4) Description of the analytic function expansion nodal (AFEN) method.

(5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods.

(6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis.

(7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

권호기사

권호기사 목록 테이블로 기사명, 저자명, 페이지, 원문, 기사목차 순으로 되어있습니다.
기사명 저자명 페이지 원문 목차
Reactor physics challenges in GEN-Ⅳ reactor design Michael J. Driscoll ;Pavel Hejzlar pp.1-10

Status and perspective of nuclear data production, evaluation and validation A. Trkov pp.11-24

Fundamentals and recent developments of reactor physics methods Nam Zin Cho pp.25-78

Evolution of nuclear fuel management and reactor operational aid tools Paul J. Turinsky ;Paul M. Keller ;Hany S. Abdel-Khalik pp.79-90

Core design for heterogeneous thorium fuel assemblies for PWR(Ⅰ) :nuclear design and fuel cycle economy Kang-mok Bae ;Myung-hyun Kim pp.91-100

Neutron cross section data library for Pd-105, Ag-109, Xe-131 and Cs-133 Yong Deok Lee ;Jong Hwa Chang pp.101-108

A study on an assessment method for improving technical specifications using system dynamics Kyung Min Kang ;Moosung Jae pp.109-117

A validation method for emergency operating procedures of nuclear power plants based on dynamic multi-level flow modeling Wei Qin ;Poong Hyun Seong pp.118-126

참고문헌 (76건) : 자료제공( 네이버학술정보 )

참고문헌 목록에 대한 테이블로 번호, 참고문헌, 국회도서관 소장유무로 구성되어 있습니다.
번호 참고문헌 국회도서관 소장유무
1 (1983.) Methods of Steady-StateReactor Physics in Nuclear Design, Academic Press 미소장
2 (1986) Biases in the Estimation ofand Its Error by Monte Carlo Methods, 미소장
3 (1994) Biases in Monte CarloEigenvalue Calculations, 미소장
4 (1997) Error Estimates andTheir Biases in Monte Carlo Eigenvalue Calculations, 미소장
5 (2003) AlternativeImplementations of the Monte Carlo Power Method, 미소장
6 (2004) Variational VarianceReduction for Monte Carlo Eigenvalue and EigenfunctionProblems, 미소장
7 (1975.) The MIT Press, 미소장
8 (1997.) Computational Methods in Engineering andScience, John Wiley & Sons 미소장
9 (1966.) Iterative Solution of Elliptic Systems, E.L. Wachspress Prentice-Hall 미소장
10 (1999) An Efficient Alternative to Solvethe Neutron Diffusion Equation, 미소장
11 (2002) Fast Iterative Methods forDiscrete-Ordinates Particle Transport Calculations, 미소장
12 (19781978) A New approach to Homogenization and GroupCondensation, 미소장
13 (1980) Spatial Homogenization Methods for LightWater Reactor Analysis, 미소장
14 (1977) InterfaceCurrent Techniques for Multidimensional ReactorCalculations, 미소장
15 (1979) An Analytic Nodal Method for Solving the Two-Group Massachusetts Institute of Technology, 미소장
16 (1995) Conformal Mapping andHexagonal Nodal Methods-I: Mathematical Foundation, 미소장
17 (1994) A New Approach of Analytic BasisFunction Expansion to Neutron Diffusion Calculation, 미소장
18 (1995) Analytic Function Expansion NodalMethod for Hexagonal Geometry, 미소장
19 (1997) Extension of AnalyticFunction Expansion Nodal Method to Multigroup Problems inHexagonal-Z Geometry, 미소장
20 (2001) The Analytic FunctionExpansion Nodal Method Refined with Transverse GradientBasis Functions and Interface Flux Moments, 미소장
21 (2002) Kinetics Calculation Under Space-Dependent Feedback in Analytic Function Expansion NodalMethod via Solution Decomposition and Galerkin Scheme, 미소장
22 (2003) Method with Half-Interface Averaged Fluxesin Mixed Geometry Nodes for Analysis of Pebble BedModular Reactor , 미소장
23 (2005.06) Improved Features and Verification of anAFEN Method Code in Hexagonal-Z 3-D Geometry forNeutron Diffusion Calculation, 미소장
24 (2002.03) User Manual for the PARCS-v2.3 BetaKinetics Core Simulator Module, Purdue University 미소장
25 (April1995) Intranodal Burnup GradientCorrection in Analytic Function Expansion Nodal Proceedings of the International Conference onMathematics and Computations, 미소장
26 (1995) Hybrid of AFEN and PEN Methodsfor Multigroup Diffusion Nodal Calculation, 미소장
27 (2001) Unified Formulation of NodalExpansion Method and Analytic Nodal Method Solutions toTwo-Group Diffusion Equations, 미소장
28 (2002) Unified Nodal Method Formulationfor Analytic Function Expansion Nodal Method Solution toTwo-Group Diffusion Equations in Rectangular Geometry, 미소장
29 (1983) Nodal Method Storage Reduction by NonlinearIteration, 미소장
30 (1989) Proc. Topical Meeting on Advancesin Nuclear Engineering Computations and RadiationShielding, 미소장
31 (1999) Acceleration of the Analytic Function Expansion NodalMethod by Two-Factor Two-Node Nonlinear Iteration, 미소장
32 (1999) Acceleration of Three-DimensionalAFEN Nodal Codes via Coarse Group Rebalance and DirectMatrix Inverse, 미소장
33 (1972) A Characteristics Formulation of the NeutronTransport Equation in Complicated Geometries, 미소장
34 (1980) A Characteristics Solutions to theNeutron Transport Equations in Complicated Geometries, 미소장
35 (1995) Optimal Polar Angles andWeights, 미소장
36 (1981) ParticleTransport Calculations with the Method of Streaming Rays, 미소장
37 (1998) A Code for Rectangular andHexagonal Lattices Based on the Method of Characteristics, 미소장
38 (2000) Whole-Core Heterogeneous TransportCalculations and Their Comparison with Diffusion Results, 미소장
39 (1999.) The Characteristics TransportCalculation for a Multi-Assembly System using Neutron PathLinking Technique, 미소장
40 (2000.) Acceleration andParallelization of the Method of Characteristics for Lattice andWhole-Core Heterogeneous Calculations, 미소장
41 (2000.) Whole-Core Neutron TransportCalculations Without Fuel-Coolant Homogenization , 미소장
42 (2002) Fusion of Method ofCharacteristics and Nodal Method for 3-D Whole-Core Transport Calculation, 미소장
43 (2002) Refinement of the 2-D/1-D Fusion Methodfor 3-D Whole-Core Transport Calculation, 미소장
44 (2002) Three-Dimensional Heterogeneous WholeCore Transport Calculation Employing Planar MOCSolutions, 미소장
45 (m&c200320032003) Diffusion-Like 3-D HeterogeneousCore Calculation with 2-D Characteristics TransportCorrection by Non-Linear Iteration Technique, 미소장
46 (2003.) Expert Group on 3-D Radiation TransportBenchmarks ? , Benchmark Specification for DeterministicMOX Fuel Assembly Transport Calculations without SpatialHomoginisation 미소장
47 (1998.) Impossibility of Unconditional Stability andRobustness of Diffusive Acceleration Schemes, 미소장
48 (1999.) Effect of MaterialHeterogeneity on the Performance of DSA for Even-Parity SNMethods International Conference on Mathematics andComputation, 미소장
49 (2003) Krylov Subspace Iterations for the Calculationof k-Eigenvalues with Sn Transport Codes, Int. Conf.Nuclear Mathematical and Computational Sciences (M&C2003) Gatlinburg, 미소장
50 (2000) CASMO CharacteristicsMethod for Two-Dimensional PWR and BWR CoreCalculation, 미소장
51 (20022002) LWR CoreCalculations with CASMO-4E Proc. Int. Conf. NewFrontiers of Nuclear Technology, 미소장
52 (2002) Cell-Based CMFD Formulation forAcceleration of Whole-Core Method of CharacteristicsCalculations, 미소장
53 (2002) Dynamic Implementation of the EquivalenceTheory in the Heterogeneous Whole Core TransportCalculation Proc. Int. Conf. New Frontiers of NuclearTechnology, 미소장
54 (2003) A Comparison of Coarse MeshRebalance and Coarse Mesh Finite Difference Accelerationsfor the Neutron Transport Calculations, Int. Conf. NuclearMathematical and Computational Sciences (M&C 2003)Gatlinburg 미소장
55 (2003) Partial Current-Based CMFDAcceleration of the 2D/1D Fusion method for 3D Whole-CoreTransport Calculations, 미소장
56 (1997) A Rebalance Approach toNonlinear Iteration for Solving the Neutron TransportEquations, 미소장
57 (1999) Convergence Analysis of theAngular-Dependent Rebalance Iteration Method in X-YGeometry, 미소장
58 (2004) Coarse-Mesh Angular DependentRebalance Acceleration of the Discrete Ordinate TransportCalculations, 미소장
59 (2004.10) The MOC Neutron TransportCalculations Accelerated by Coarse-Mesh Angular DependentRebalance, 미소장
60 (20022002) A Time-Dependent NeutronTransport Code Coupled with the Thermal-Hydraulics CodeATHLET Proc. Int. Conf. New Frontiers of NuclearTechnology, 미소장
61 (1958) The Application of Reactor Kinetics to theAnalysis of Experiments, 미소장
62 (1966) Quasistatic Treatment of Spatial Phenomena inReactor Dynamics, 미소장
63 (1969) Accuracy of the QuasistaticTreatment of Spatial Reactor Kinetics, 미소장
64 (1970) Alternating Direction Methodsfor the Reactor Kinetics Equations, 미소장
65 (1996) Diffusion Theory Methods forSpatial Kinetics Calculations, 미소장
66 (1992) Nonlinear Iteration Strategy for NEM, 미소장
67 (2003) Using the OECD/NRCPressurized Water Reactor Main Steam Line BreakBenchmark to Study Current Numerical and ComputationalIssues of Coupled Calculations, 미소장
68 Nuclear Energy Agency Committee on ReactorPhysics, H.B. Finnemann and A. Galati 미소장
69 (apr.1999) PWR MSLB Benchmark. Volume 1 :Final Specifications, NEA/NSC/DOC (99)8, Nuclear EnergyAgency/Nuclear Science Committee 미소장
70 (2002) VVER-1000 Coolant Transient Benchmark ?PHASE I (V1000CT-1) Volume I : Final Specifications, 미소장
71 (1999) Dimensionally Adaptive Neutron Kinetics for Multidimensional Reactor Safety Transients ? 1 NewFeatures of RELAP5/PANBOX, 미소장
72 (1999) Dimensionally Adaptive Neutron Kinetics forMultidimensional Reactor Safety Transients ? II Dimensionally Adaptive Switching Algorithms, 미소장
73 (2003) Sensitivity Studies for Main SteamLine Break Exercises 2 and 3 with RELAP5/PANBOX, 미소장
74 (2002) Development of a Newton-Krylov Solver in FORMOSA-B, 미소장
75 Implementation of a Newton-BiCGSTAB Solver to Treat the Strong Non-Linearity in theFORMOSA-B Boiling Water Reactor Core Simulator Code, 미소장
76 (2003) Nuclear Mathematical and Computational Sciences(M&C 2003) , Gatlinburg, USA, 미소장

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