• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.608-616
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• 2015

The migration of silver (Ag) in silicon carbide (SiC) and $^{110m}Ag$ through SiC of irradiated tristructural isotropic (TRISO) fuel has been studied for the past three to four decades. However, there is no satisfactory explanation for the transport mechanism of Ag in SiC. In this work, the diffusion coefficients of Ag measured and/or estimated in previous studies were reviewed, and then pre-exponential factors and activation energies from the previous experiments were evaluated using Arrhenius equation. The activation energy is $247.4kJ{\cdot}mol^{-1}$ from Ag paste experiments between two SiC layers produced using fluidized-bed chemical vapor deposition (FBCVD), $125.3kJ{\cdot}mol^{-1}$ from integral release experiments (annealing of irradiated TRISO fuel), $121.8kJ{\cdot}mol^{-1}$ from fractional Ag release during irradiation of TRISO fuel in high flux reactor (HFR), and $274.8kJ{\cdot}mol^{-1}$ from Ag ion implantation experiments, respectively. The activation energy from ion implantation experiments is greater than that from Ag paste, fractional release and integral release, and the activation energy from Ag paste experiments is approximately two times greater than that from integral release experiments and fractional Ag release during the irradiation of TRISO fuel in HFR. The pre-exponential factors are also very different depending on the experimental methods and estimation. From a comparison of the pre-exponential factors and activation energies, it can be analogized that the diffusion mechanism of Ag using ion implantation experiment is different from other experiments, such as a Ag paste experiment, integral release experiments, and heating experiments after irradiating TRISO fuel in HFR. However, the results of this work do not support the long held assumption that Ag release from FBCVD-SiC, used for the coating layer in TRISO fuel, is dominated by grain boundary diffusion. In order to understand in detail the transport mechanism of Ag through the coating layer, FBCVD-SiC in TRISO fuel, a microstructural change caused by neutron irradiation during operation has to be fully considered.

• Nuclear Engineering and Technology
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• v.3 no.3
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• pp.121-128
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• 1971

A unified modal-nodal expansion of tile angular distribution of neutron flux in one spatial dimension is considered, following the proposal of Harms. Several standard nodal and/or modal methods of analysis are shown to be specializations of this technique. The modal-nodal moment from of the mono-energetic transport equation with isotropic sources and scattering is derived and the infinite medium eigenvalue problem solved. The technique is shown to yield results which approximate the exact value of the inverse diffusion length in non-multiplying media more accurately than standard methods of equal or somewhat greater computational complexity.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.27-32
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• 1997

In this paper, a new multigrid method is developed to solve the reactor eigenvalue problems. The new algorithm can be used in any matrix equation concerned with the eigenvalue problem. The finite difference neutron diffusion problem is considered demonstration of the performance of the new multigrid algorithm. The numerical results show that the new multigrid algorithm works well and requires much shorter (7~10 times) computing time compaired to the production code VENTURE.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.33 no.3
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• pp.91-99
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• 1984

Time-optimal control for dmping a one-dimensional xenon-induced spatial power oscillations in pressurized water reactor is studied. Linearized system equations describing the spatial xenon oscillations have been derived based on lambda mode analysis. Optimal control strategies, eventually bang-bang controls, have been drawn applying Pontryagins Minimum Principle, subject to a band constraint on available contros strength. Validity of the linearized system equations and optimal control strategies derived has been demonstrated through conputer simulations which incorporate the finite difference method for one dimensional axial geometry, for the soulution of the two-group neutron diffusion equations. The results obtained through computer simulations show that xenon-induced transients can be suppressed successfully with bang-bang control.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.79-86
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• 1998

The nonlinear analytic function expansion nodal (AFEN) method is applied to the solution of the time-dependent neutron diffusion equation. Since the AFEN method requires both the particular solution and the homogeneous solution to the transient fixed source problem, the derivation solution method is focused on finding the particular solution efficiently. To avoid complicated particular solutions, the source distribution is approximated by quadratic polynomials and the transient source is constructed such that the error due to the quadratic approximation is minimized. In addition, this paper presents a new two-node solution scheme that is derived by imposing the constraint of current continuity at the interface corner points. The method is verified through a series of applications to the NEACRP PWR rod ejection benchmark problem.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.141-145
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• 1994

A computer code for solving the 3-D time-dependent multigroup neutron diffusion equation by a coupled reactor kinetics method recently developed has been developed and for evaluating its applicability in CANDU transient analysis applied to a 3-D kinetics benchmark problem which reveals non-uniform loss of coolant accident followed by an asymmetric insertion of shutdown devices. The performance of the method and code has been compared with the CANDU design code, CERBERUS, employing a finite difference improved quasistatic method.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.142-147
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• 1996

The analytic function expansion nodal (AFEN) method has been successfully applied to two-group neutron diffusion problems. In this paper, the AFEN method is extended to solve general multigroup equations for any type of geometries. Also, a suite of new nodal codes based on the extended AFEN theory is developed for hexagonal-z geometry and applied to several benchmark problems. Numerical results obtained attest to their accuracy and applicability to practical problems.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.21-27
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• 1996

KAERI has recently developed the nuclear design code MASTER for the application to reactor physics analyses for pressurized water reactors. Its neutronics model solves the space-time dependent neutron diffusion equations with the advanced nodal methods. The major calculation categories of MASTER consist of microscopic depletion, steady-state and transient solution, xenon dynamics, adjoint solution and pin power and burnup reconstruction. The MASTER validation analyses, which are in progress aiming to submit the Uncertainty Topical Report to KINS in the first half of 1996, include global reactivity calculations and detailed pin-by-pin power distributions as well as in-core detector reaction rate calculations. The objective of this paper is to give an overall description of the CASMO/MASTER code system whose verification results are in details presented in the separate papers.