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A conceptual design of a passive core cooling system (PCCS) was proposed in order to removal residual core heat during accident scenarios. Strong coupling feedback effect exists between natural convective heat transfer of the PCCS and the temperature distribution within the reactor core, which has a significant impact on the heat removal performance of the PCCS and the peak fuel temperature. An integral computational model dedicated to micro gas-cooled reactor, which includes both the PCCS and the reactor core, was developed. Comparisons of results predicted by this model and reference results of two benchmark problem, i.e. exercises of MHTGR and natural convection in a square cavity, show a satisfactory agreement, which thus validate the model. Thereafter, in order to explore the operational characteristics of the PCCS proposed in this work as well as the temperature response of the reactor core, and to assess its mitigative performance during accident scenario, this CFD model was then employed to analyze both the steady state under normal operation condition and the pressurized-loss-of-forced-cooling (PLOFC) accident. The calculation results show that the natural circulation of air is steady and regular during normal operation. The heat loss due to the PCCS is acceptably small, which will not lead to significant impact on the core temperature. During the PLOFC accident, the heat removal power of the PCCS is high in the early phase, whereas its performance gradually decreases as the temperature of the reactor vessel decreases in the long term. In general, the PCCS can effectively remove the residual core heat, thus maintaining the reactor integrity. Consequently, the integral CFD model created in this work can be applied to the design and the safety analysis of the PCCS for the micro gas-cooled reactor.

Accurate nuclide composition of nuclear spent fuel (NSF) assemblies is the most important condition for both criticality safety analysis and radiation safety analysis of NSF. The calculation of radioactive source terms is a typical application of nuclide composition calculation for NSF. There are drawbacks in the existing programs, such as the insufficient nuclides in the database and the inability to fully simulate the irradiation process, which limit the reliability and economy of NSF reprocessing safety analysis. Based on the completely autonomous PWR-core analysis software NECP-Bamboo, a nuclide composition calculation program named Bamboo-SFuel was independently developed for commercial PWR NSF, and the measured data of post-irradiation experiment (PIE) in SFCOMP database developed by NEA (Nuclear Energy Agency) were used to quantitatively verify and analyze the program. The experiments employed cover a variety of fuel types, component types and operating conditions including power level, boron concentration, fuel temperature and moderator temprature. In addition, a VERA benchmark problem was computed by using both Bamboo-SFuel and SCALE to verify the calculation accuracy of the source term including radioactivity, decay heat, photon source and neutron source. The influence of different depletion data libraries including 3838, 1547 and 233 nuclides respectively on the calculated nuclide composition and source term was also investigated. Encouraging conclusions are demonstrated by the numerical results. Bamboo-SFuel can accurately calculate both the nuclide composition and the radioactive source term of commercial PWR NSF under different irradiation conditions. Among the actinide nuclides, the average relative deviation of 247Cm is the largest (-31.059%), while the relative deviation of 125Sb in fission products is the largest, which is 27.35%. About the source term, the maximum relative deviation of photon source term is 0.69%, the maximum relative deviation of spontaneous fission neutron source term is 0.17%, and the maximum relative deviation of (α,n) neutron source term is 0.43%. The influence of different depletion databases in NECP-Bamboo code system on the calculation results of important nuclide components is very small. But the influence on the calculation results of the total radioactive source items like photon source term is as large as 99.99%, because the conservation of photon source terms is not taken into account when the depletion database is compressed from 1 547 to 233 nuclides, many nuclides with large photon source terms are also deleted, such as 231Th, and the photon yield of these nuclides at some energy points is large, resulting in a large deviation in the calculation of corresponding energy groups at these energy points. By using the depletion data library with 1 547 nuclides contained in the NECP-Bamboo code system, all the important nuclides concerning in both criticality safety analysis and radiation safety analysis of NSF can be accurately evaluated. The maximum relative deviation in nuclide composition for criticality safety analysis is 238Pu with -29.443%, while 0.69% in source term calculation for radiation safety analysis.

As the cornerstone of the development of the computational tool for the coupled neutronics and thermal-hydraulics analysis dedicated to the micro gas-cooled reactor, neutronic diffusion solver was developed in this work based on the generic 3D finite-element CFD code COMSOL Multiphysics by taking advantage of its equation-based modeling functionality “coefficient form PDE interface”. The steady-state and transient neutronic diffusion equations are solved using the eigenvalue solver and transient solver of this code, respectively. The neutronic diffusion solver developed in this work was firstly validated against two benchmark problems, i.e. 2D-TWIGL and 3D IAEA PWR problems respectively. The comparison of the calculation results of this solve with the reference values offered by the corresponding benchmark problems is satisfactory. Thereafter, this neutronic diffusion solve was applied to the micro gas-cooled reactor criticality analysis, with the group constants (including two-group and 25-group) generated by the Monte-Carlo code RMC. The calculation results of neutronic diffusion were also compared with the reference values derived by the Monte-Carlo transport calculations using the RMC code based on both continuous-energy-spectrum and multi-group methods. In general, the results obtained by solving both the two-group or 25-group neutronic diffusion equations, such as effective multiplication factor and three-dimension power distribution, can match well with the corresponding multi-group Monte-Carlo simulation. Whereas when comparing with the continuous-energy-spectrum MonteCarlo results, adopting the 25 energy groups is able to gain better accuracy in contrast to the two energy groups, indicating that the energy-group division has a significant impact on the neutronics for the micro gas-cooled reactor. The accuracy of prediction can be improved by adopting a more sophisticated energy group structure. However, the required computational source and time will dramatically increase with the number of energy groups.

The prismatic super-safe gas-cooled reactor is one of the advanced reactor types that can be designed as the movable micro nuclear power devices applied in some specific scenarios, such as remote areas, islands, post disaster reconstruction, border national defense and so on. The reactor has superior inherent safety because of graphite core with high temperature resistance and large heat capacity, ceramic particle dispersion fuel almost completely containing fission products, the low power and power density. What’s more, on the basis of high temperature gascooled reactor, the SiC matrix and nofuel zone of fuel pellet further enhances the barrier effect on fission products. Through the design of burnable poison and negative temperature feedback, the reactor can realize automatic shutdown only relying on negative reactivity caused by temperature rise, even if all control rods are pulled out without any emergency measures under accident conditions. From a physical point of view, the risk of core melting and massive release of radioactive materials is avoided. To analyze its core physical characteristics, a core loading scheme and a reactivity control strategy were proposed, and then burnup, distribution of power, distribution of neutron flux density, neutron energy spectrum, reactivity temperature coefficient, xenon oscillation, and xenon reactivity were investigated based on the core model by using Monte Carlo procedure. The results show that the requirement of 5 MW thermal power and 3 a lifetime can be met and the average discharge burnup is 18 700 MW·d/tU. If the fuel enrichment increases from 85% to 15%, the lifetime will increase to 10 a and the maximum discharge burnup is 79 600 MW·d/tU. The radial distribution of power is uniform, while the axial is in the shape of piecewise concave curve similar to the distribution trend of thermal neutrons. The maximum neutron flux density is as low as 43×1013 cm-2·s-1 at the beginning of lifetime, and the neutron energy spectrum is greatly affected by temperature but less by burnup. Reactivity temperature coefficients, including negative fuel temperature coefficient, negative core graphite temperature coefficient, and small positive reflector graphite temperature coefficient, are influenced by burnup and temperature significantly, which makes the core have the safety of automatic shutdown only relying on negative temperature feedback under accident conditions. The amplitude of xenon oscillation is very small and the reactivity of iodine well is only -110 pcm when the reactor is shutdown at full power, thus the core has great stability. These results will instruct the followup development of gascooled reactor micro devices.

Analyses of representative accident scenarios are imperative to assess and evaluate the inherent safety features of a prismatic modular high temperature gascooled reactor (HTGR). The accident progression of a HTGR is typically characterized by the coupling of multiphysics phenomena. However, computational codes available for the HTGR accident analyses are still relatively rare at present, especially those dedicated to the prismatictype core design. In addition, many of these codes are either lumpedparameter that cannot give the detailed threedimension (3D) distributions of key variables (e.g. fuel temperature), or are incapable of capturing the coupling of major phenomena. As a consequence, a generic CFD code, i.e. COMSOL multiphysics, was employed in this work to create a detailed 3D geometry representing the reactor, including fuel assemblies containing fuel and coolant channels, graphite moderator blocks, side reflector as well as the reactor pressure vessel. Specifically, the passive residual heat removal system (RCCS) was simply treated as a convective boundary condition due to the lack of detailed design scheme right now. A point kinetics module was prepared in this work using C++ language to account for the transient neutron kinetics. The Gear algorithm was applied to numerically solve the system of equations of the point kinetics, given that the stiffness is relatively large. The C++ was then compiled into dynamic link library (DLL), which was accessed by the COMSOL simulation in the form of an external function, with the fuel and moderator temperatures evaluated by COMSOL as input arguments. The steady state under normal operation condition was first calculated, which served as the initial condition for the subsequent transient analyses for three selected typical accidents, including pressurized loss of forced cooling (PLOFC), PLOFC without scram (PLOFC+ATWS), and reactivity insertion without scram (RIA+ATWS) accidents. Finally, sensitivity analysis was carried out on the effect of radiative emissivity between reactor pressure wall and RCCS. The simulation results show that the peak fuel temperature throughout the whole accident progression is remained far below its upper limit value (1 620 ℃) with a considerably large margin, implying that the integrity of fuel assemblies can be successfully guaranteed under the accident scenarios concerned in the present work. Moreover, enhancement of the RCCS heat removal capacity can significantly relief the accident consequence for PLOFC accident but lead to an inverse trend for ATWS accident, which underlines the importance of further comprehensive analysis and modeling validation. To conclude, the computational model developed in this work along with the derived results can serve as the starting point for future development of the accident analysis tool, which is applicable for the incoming indepth safety analysis and assessment works.