Translated Abstract
Molten Salt Reactor (MSR) is the only one using fluid-fuel in Generation IV advanced nuclear reactor systems. On one hand the fuel salt of MSR serves as a fuel and a coolant simultaneously and the energy produced by fission reaction is released in fuel salt directly, on the other hand the fuel salt circulates in the whole primary loop. These properties lead MSR to be a new reactor technology different from traditional solid fuel reactor and bring challenges in the field of neutronics and thermal-hydraulics calculation. Therefore, neutronics/thermal-hydraulics coupled method and steady and transient analysis of MSR have academic signification and practical value for MSR design and optimization.
Firstly, neutronics model and multi-channel thermal hydraulics model applicable for MSR are developed based on characteristic of fuel circulation and geometry of channel-type MSR, and assembly homogenization method, neutron diffusion equation calculation method, steady-state thermal-hydraulics calculation method and neutronics/thermal-hydraulics coupling method are studied. 1) Suitable assembly homogenization method is developed based on the geometry characteristics of MSRE and TMSR-LF. The study shows that eigenvalue error is minimum by applying 14 groups and 7 groups few group constant for MSRE and TMSR-LF, respectively. 2) Quasi-diffusion equation is derived by considering anisotropy of MSR when control rod exists. The result indicate that quasi-diffusion equation is more accurate than traditional diffusion equation in case of strong anisotropic problem and their calculation efficiency are in the same level. 3) Considering fuel flow, suitable iteration scheme for steady neutronics is build, and the flow is distributed for parallel-channels based on equal pressure drop. 4) Explicit coupling is performed in coupling calculation between neutronics and thermal-hydraulics. The least square method is employed in the cross-section feedback and three-order expansion give minimum error of fitting cross section. Coupled neutronics/thermal-hydraulics steady code for MSR is developed based on above method.
Secondly, the diffusion-based transient method of MSR is studied, including neutron kinetic method, transient method of multi-channel thermal-hydraulics and coupling method. 1) The prediction-correction improved quasi-static method (PCIQS) is employed to eliminate the nonlinear iteration between amplitude function and shape function and to improve efficiency. 2) Mass and momentum conservation equations are discreted by implicit difference method and coupled by equal pressure drop, and graphite heat conduction equation is solved by effective heat transfer coefficient method. 3) Transient of neutronics and thermal-hydraulics is explicitly coupled in medium time step to reduce computational load. A code MOREL for MSR analysis is developed based on above theory and numerical model and validated by some kinetic benchmark problems and MSRE zero-power experiment data. The validation results indicate that MOREL can meet the requirement of MSR analysis.
Finally, steady-state and transient analysis for TMSR-LF preliminary scheme are performed. At steady-state, 1) the reactivity coefficient of fuel salt and graphite are -9.89 pcm·K-1 and -7.76 pcm·K-1, respectively, which means the TMSR-LF has inherent safety feature. 2) keff decreases with increasing of fuel flow. The difference of reactivity at zero and rated fuel flow is 169 pcm. 3) keff decreases with increasing external loop length and tends to stable after about 80 s. 4) The fuel flow rate has a significant influence on the distribution of DNP with small decay constant. 5) The heat source in graphite and convective heat transfer coefficient between salt and graphite have a significant influence on the graphite temperature. At transient, 1) the core power changes oscillatorily in zero-power pump start-up case. Positive reactivity is introduced in pump coast-down case, and power increase and then drop down due to the temperature feedback. 2) TMSR-LF core power varies more rapidly with larger fuel flow rate at given equal reactivity introduction. After introduction of positive 566.5 pcm, delayed criticality and prompt supercriticality happen for zero fuel flow rate and rated fuel flow, respectively. 3) Overheating of fuel inlet temperature introduces negative reactivity, which makes reactor remain in safety state. Supercriticality appears when fuel inlet temperature is subcooling 70 K. 4) Transient behavior of TMSR-LF caused by local parameter is studied and results show that TMSR-LF has a good safety feature.
In short, research on steady and transient method of neutronics/thermal-hydraulics are performed for the channel-type MSR. A code named MOREL for steady and transient analysis of MSR with coupling neutronics and thermal-hydraulics is developed. MOREL is validated by some benchmark problems and is used to analyze the steady and transient characteristic for TMSR-LF proposed by China. The conclusions obtained provide reference for revelation of coupling mechanism of neutronics and thermal-hydraulics, transient safety characteristics and accident development and the MOREL code can be also used for design optimization and safety analysis of MSR.
Translated Keyword
[Coupling of neutronics and thermal-hydraulics, Molten salt reactor, Neutron diffusion, Prediction-correction improved quasi-static method, Quasi diffusion, Transient]
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