A New Numerical Nuclear Reactor Neutronics Code SHARK

In order to establish the next-generation reactor physics calculation method based on the numerical nuclear reactor technology and realize high-fidelity modeling and calculation, a new numerical nuclear reactor neutronics code SHARK is developed. The code is based on the direct transport method with construct solid geometry (CSG) method, advanced subgroup resonance method, direct transport MOC method in rectangle and hexagonal geometry, large-scale parallel, and CMFD acceleration method. The C5G7, macro BEAVRS and VERA benchmarks are verified to show the accuracy of the code and method. Numerical results show good accuracy and calculation performance of SHARK, and the direct transport method can be adopted on numerical nuclear reactor calculation.

The Differences Between the Two Forms of Semi-Analytical Nodal Methods on Solving the Third-Order Simplified Spherical Harmonics Method Equations1

Solving the third-order simplified spherical harmonics method (SP3) equations is one of the key points in the development of advanced reactor calculation methods and has been widely concerned. The semi-analytical nodal method (SANM), based on transverse-integrated diffusion equation, has the advantages of high accuracy and convenience for multigroup calculation. Due to its advantages, the method is expected to be used in solving the SP3 equations. However, the traditional SANM is not rigorous since the expansion process does not take the special modality of the SP3 equations and their analytical solutions into consideration. There are two modalities of the SP3 equations, so there are two traditional SANM forms on solving the SP3 equations, and the differences between the two forms will be very important in further research on the SANM. A code is developed to solve the SP3 equations under the two different forms. After the calculation of the same benchmark, the difference between the two forms on solving the SP3 equations is found. According to the results, and in view of the special modality of the SP3 equations, points on a more rigorous SANM for solving the SP3 equations are discussed.

The Generation of Few-Group Constants for Fast Reactor Analysis

A way to generate the few-group cross sections for fast reactor calculation is presented in this paper. It is based on the three steps computational scheme. In the first step, the ultrafine method is used to solve the slowing down equation based on the ultrafine group cross section generated by NJOY. Optional 0D or 1D calculation is used to collapse energy group into broad energy groups. In the second step, the 2D RZ calculation using SN method is performed to obtain the space dependent neutron spectra to collapse broad energy groups into few groups. The anisotropic scattering is well handled by the direct SN calculation. Finally, the full core calculation is performed by using the 3D SN nodal method. The results are compared with continuous energy Monte-Carlo calculation. Both the cross section generated in the first step and the final keff in the last step are compared. The results match well between the three steps calculation and Monte-Carlo calculation.

CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.