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FUNDAMENTALS AND RECENT DEVELOPMENTS OF REACTOR PHYSICS METHODS
CHO NAM ZIN Korean Nuclear Society 2005 Nuclear Engineering and Technology Vol.37 No.1
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 multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (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.
Cho, Nam Zin,Jo, YuGwon,Yuk, Seungsu Elsevier 2017 Annals of nuclear energy Vol.110 No.-
<P><B>Abstract</B></P> <P>This work presents a new consistent derivation of the multigroup transport equations from the original continuous-energy transport equation, achieving group-wise condensation equivalence. The derivation of the multigroup transport equations currently in use involves an important approximation in the total cross section. The homogeneity and isotropy restoration (HIRE) theory described in this paper removes the angle dependency in the group condensed total cross section of the multigroup transport equations. The HIRE theory also restores the homogeneity of each material region, and it frees us from the higher-order moment scattering cross sections. To preserve the reaction rate of the original continuous-energy transport equation, the partial current discontinuity factor (PCDF) is introduced, that is instrumental in the theory. The theory was tested on several pin-cell problems, and the numerical results show that the new multigroup transport equations successfully reproduced the properties of the original continuous-energy transport equation.</P>
SOME OUTSTANDING PROBLEMS IN NEUTRON TRANSPORT COMPUTATION
Cho, Nam-Zin,Chang, Jong-Hwa Korean Nuclear Society 2009 Nuclear Engineering and Technology Vol.41 No.4
This article provides selects of outstanding problems in computational neutron transport, with some suggested approaches thereto, as follows: i) ray effect in discrete ordinates method, ii) diffusion synthetic acceleration in strongly heterogeneous problems, iii) method of characteristics extension to three-dimensional geometry, iv) fission source and $k_{eff}$ convergence in Monte Carlo, v) depletion in Monte Carlo, vi) nuclear data evaluation, and vii) uncertainty estimation, including covariance data.
Mathematical Adjoint Solution to Analytic Function Expansion Nodal (AFEN) Method
Cho, Nam-Zin,Hong, Ser-Gi Korean Nuclear Society 1995 Nuclear Engineering and Technology Vol.27 No.3
The mathematical adjoint solution of the Analytic Function Expansion (AFEN) method is found by solving the transposed matrix equation of AFEN nodal equation with only minor modification to the forward solution code AFEN. The perturbation calculations are then performed to estimate the change of reactivity by using the mathematical adjoint The adjoint calculational scheme in this study does not require the knowledge of the physical adjoint or the eigenvalue of the forward equation. Using the adjoint solutions, the exact and first-order perturbation calculations are peformed for the well-known benchmark problems (i.e., IAEA-2D benchmark problem and EPRI-9R benchmark problem). The results show that the mathematical adjoint flux calculated in the code is the correct adjoint solution of the AFEN method.
ON SOME OUTSTANDING PROBLEMS IN NUCLEAR REACTOR ANALYSIS
Cho, Nam-Zin Korean Nuclear Society 2012 Nuclear Engineering and Technology Vol.44 No.2
This article discusses selects of some outstanding problems in nuclear reactor analysis, with proposed approaches thereto and numerical test results, as follows: i) multi-group approximation in the transport equation, ii) homogenization based on isolated single-assembly calculation, and iii) critical spectrum in Monte Carlo depletion.