Jin-Young Cho

Whole Core Transport Calculation Employing Hexagonal Modular Ray Tracing and CMFD Formulation

A whole core transport module for hexagonal cores is developed and implemented in the DeCART code. The module consists of a whole core ray tracing kernel for solving the two-dimensional (2-D) method of characteristics (MOC) equation and a coarse-mesh finite difference (CMFD) kernel to accelerate the MOC transport iteration. Whole core ray tracing is realized by incorporating a hexagonal-assembly-based modular ray tracing scheme. The complete path linking constraint forthe modular ray is achieved by adjusting the ray angle and the ray spacing in the range of [0, 30°], and the complete reflection constraint at the problem boundary is satisfied by defining the corresponding reflection angles at the reflection surfaces. The hexagonal CMFD kernel employs unstructured nodes that can treat the irregular-shaped gap cells as well as the regular hexagonal cells. Some features such as cell ray approximation and modified cycle ray scheme are employed to reduce the memory requirements for the segment information and the boundary angular fluxes, respectively. The solution accuracy and execution performance of the hexagonal module are examined for the C5G7 hexagonal variation problems that are established by modifying the original C5G7MOX 3-D extension benchmark. The CMFD kernel shows a significant speedup of 60 in the 2-D core problem. The cell ray approximation does not violate the original solution accuracy when using the default ray spacing of the DeCART code. The modified cycle ray scheme shows its superiority over the simple core ray sweeping scheme in terms of the memory requirement and the original cycle ray scheme in terms of the computing time. Compared with the Monte Carlo solutions, the DeCART solution agrees to within 40 pcm for the eigenvalue and 2% for the pin power distribution.

Axial SPN and Radial MOC Coupled Whole Core Transport Calculation

The Simplified PN (SPN) method is applied to the axial solution of the two-dimensional (2-D) method of characteristics (MOC) solution based whole core transport calculation. A sub-plane scheme and the nodal expansion method (NEM) are employed for the solution of the one-dimensional (1-D) SPN equations involving a radial transverse leakage. The SPN solver replaces the axial diffusion solver of the DeCART direct whole core transport code to provide more accurate, transport theory based axial solutions. In the sub-plane scheme, the radial equivalent homogenization parameters generated by the local MOC for a thick plane are assigned to the multiple finer planes in the subsequent global three-dimensional (3-D) coarse mesh finite difference (CMFD) calculation in which the NEM is employed for the axial solution. The sub-plane scheme induces a much less nodal error while having little impact on the axial leakage representation of the radial MOC calculation. The performance of the sub-plane scheme and SPN nodal transport solver is examined by solving a set of demonstrative problems and the C5G7MOX 3-D extension benchmark problems. It is shown in the demonstrative problems that the nodal error reaching upto 1,400 pcm in a rodded case is reduced to 10pcm by introducing 10 sub-planes per MOC plane and the transport error is reduced from about 150pcm to 10pcm by using SP3. Also it is observed, in the C5G7MOX rodded configuration B problem, that the eigenvalues and pin power errors of 180 pcm and 2.2% of the 10 sub-planes diffusion case are reduced to 40 pcm and 1.4%, respectively, for SP3 with only about a 15% increase in the computing time. It is shown that the SP5 case gives very similar results to the SP3 case.

DeCART Depletion Calculation for PMR200 Two-Dimensional Core

This paper is to evaluate the depletion capability of the DeCART code for PMR200 two-dimensional core. DeCART solves the burnup equation by the matrix exponential based on the Krylov Subspace method. The depletion calculation is performed up to 270 EFPD by using total 14 burnup steps. The double heterogeneity effect is resolved by introducing the RPT method. The DeCART solutions by using a 47-G PWR neutron library are compared with the McCARD Monte Carlo solution which uses ENDF/B-VII. DeCART shows about maximum 1200 pcm eigenvalue, about maximum 2.5% block and 6.0% pin power differences near 180 EFPDs. The differences are mainly due to the library and the geometrical model discrepancy. While the reference calculation is performed by imposing the vacuum condition for the vessel outside, DeCART uses a zigzag-type boundary model. The use of the VHTR neutron library that is scheduled to develop reduces the eigenvalue differences. In the computing time, DeCART requires about 2 and half hours for 14 depletion steps. Therefore, it is concluded that the DeCART code produces a reasonable result for the VHTR core depletion calculation within an affordable computing time.Copyright

Lumped-Refined Multichannel Calculation Scheme for a High-Fidelity Thermal-Hydraulic Analysis by a Neutronics Code Coupling

A lumped-refined multichannel analysis scheme is developed for a high-fidelity thermal-hydraulic (T-H) calculation through neutronics code coupling and applied to a control element assembly (CEA) ejection accident of the Ulchin Unit 3 nuclear power plant to quantify the conservatism of the conventional scheme. The high-fidelity core minimum departure from nucleate boiling (DNB) ratio calculation is realized by coupling more than two TORC dynamic link libraries (DLLs) under the control of the neutronics code, one for the lumped multichannel calculation and the others for the refined subchannel calculations. Realistic radial boundary conditions are supplied from the lumped multichannel calculation to the refined TORC DLL through the neutronics code. The CEA ejection accident problem is simulated from the DNB limiting conditions for operation condition, which is searched by adjusting the core radial peaking factor at a 30% axial offset power shape. The results indicate that the simplified hot-channel model contains ~15 and 5% conservatism in the core minimum DNB ratio and in the number of failed fuel rods, respectively, and reveals that those conservatisms are mainly due to the unrealistic isolated boundary condition. Therefore, it is concluded that the developed scheme can be effectively used to quantify the conservatism of a conventional DNB evaluation scheme.