Development of a subchannel code for blockage accidents of LMFRs based on the 3D fuel rod model

To predict the thermal-hydraulic(T/H) parameters of the reactor core for liquid-metal-cooled fast reactors(LMFRs), especially under flow blockage accidents, we developed a subchannel code called KMC-FB.This code uses a time-dependent, four-equation, singlephase flow model together with a 3D heat conduction model for the fuel rods, which is solved by numerical methods based on the finite difference method with a staggered mesh. Owing to the local effect of the blockage on the flow field, low axia...     

Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2 and 8×8-4 UO2 Fuel Assemblies

The measured pellet average inventories of actinides and fission product nuclides on the fifteen samples taken from a three-cycle irradiation BWR 8×8-2 UO₂ assembly were compared with those of assembly burnup calculations using a collision probability method (SRAC) with the JENDL-3.2 nuclear data library. The present calculations overestimate the inventories of ²³⁵U, well reproduce those of ²³⁹Pu and ²⁴⁰Pu, yet underestimate those of ²³⁶U, ²³⁷Nd, ²³⁸Pu, ²⁴¹Pu, and ²⁴²Pu. The inventories of minor actinides are underestimated by the present analysis except for ²⁴¹Am. The major FP nuclides contributing to neutron absorption such as Nd, Cs, Eu, and Sm are almost well reproduced by the present calculations. The measured pellet average burnups and major actinide inventories on the twenty samples taken from four BWR 8×8-4 UO₂ assemblies were also compared with those of the burnup calculations using SRAC and a continuous energy Monte Carlo burnup analysis code (MVP-BURN). Most of the calculated pellet average burnups of both codes agree with the measurements within the range of ±10%. The general trends of the measured pellet radial distributions of actinide and FP nuclides on six samples of the 8×8-4 UO₂ assemblies were well reproduced by the burnup calculations of MVP-BURN.     

Construction of fault diagnosis system for control rod drive mechanism based on knowledge graph and Bayesian inference

Knowledge graph technology has distinct advantages in terms of fault diagnosis. In this study, the control rod drive mechanism(CRDM) of the liquid fuel thorium molten salt reactor(TMSR-LF1) was taken as the research object, and a fault diagnosis system was proposed based on knowledge graph. The subject–relation–object triples are defined based on CRDM unstructured data, including design specification, operation and maintenance manual, alarm list, and other forms of expert experience. In this stu...     

Creep damage characterization of UNS N10003 alloy based on a numerical simulation using the Norton creep law and Kachanov–Rabotnov creep damage model

The calculation of inelastic creep damage is important for the structural integrity evaluation of the elevated temperature structure in a thorium molten salt reactor(TMSR). However, a creep damage theory model and numerical simulation method have not been proposed for the key materials(UNS N10003 alloy) in the TMSR. In this study, creep damage characterization of UNS N10003 alloy is investigated using the Norton creep law and Kachanov–Rabotnov(K–R) creep damage model. First, the creep experimental data of the UNS N10003 alloy at 650 °C were adopted to fit the material constants of the two models. Then, the creep damage behavior of the UNS N10003 alloy was analyzed and discussed under uniaxial and multi-axial stress states. The results indicated that the K–R creep damage model is more suitable for the UNS N10003 alloy than the Norton model. Finally, the numerical simulation method was developed by a user-defined UMAT subroutine and subsequently verified through a finite element analysis(FEA). The FEA results were in agreement with the theoretical solutions. This study provides an effective method for the inelastic creep damage analysis of the elevated temperature structure in the TMSR.