Chung-Chan Lee

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.

Ex-Core Detector Response Evaluation of the SMART Reactor by Using the DORT Code

The development of a 660 MWt integral reactor SMART (System integrated Modular Advanced ReacTor), called SMART-660, has been initiated in Korea. A preliminary evaluation of the detector response functions was performed for the ex-core detector of the SMART-660. First, a candidate position to install the ex-core detector was selected by considering the reactor configuration, the detector performance, and the thermal neutron flux distribution in the reactor assembly. Second, the detector response functions such as the shape annealing function (SAF) and assembly weighting factor (AWF) for the candidate ex-core detector were calculated by using an adjoint transport calculation with the DORT code and the BUGLE-96 cross-section library. The GEOSHIELD program was used to generate the DORT input file and to process the DORT output file with a graphic visualization. This study has shown that the response functions of the ex-core detector are only sensitive to some parts of the reactor and additional efforts are required to reduce the locality of the ex-core detector response.

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.

Radiation Shielding Analysis for the Reactor Assembly of the SMART Reactor

A 660 MWt integral reactor SMART (System integrated Modular Advanced ReacTor), called SMART-660, is under development in Korea as a national research and development project to supply energy for a seawater desalination as well as an electricity generation. A radiation shielding analysis has been carried out to evaluate the shielding design of the SMART-660. Two-dimensional discrete ordinates transport code DORT was used to calculate the neutron and gamma dose rates and the fast neutron fluences in the reactor pressure vessel assembly (RPV). The reactor assembly of the SMART-660 was modeled as an R-Z geometry through an azimuthal homogenization of the reactor components. The GEOSHIELD program was used to generate the DORT input file and process the DORT output file with a graphic visualization. The neutron 47-group and gamma 20-group BUGLE-96 cross-section library was used for the DORT calculation. It is found that the integrity of the RPV is preserved throughout the reactor lifetime and the shielding system of the SMART-660 is satisfactory to prevent radiation escaping from the reactor core.