Translated Abstract
Fluoride-Salt-Cooled, High-Temperature Reactor (FHR) is one of generation IV small modular reactor, which employs both the fluoride salt as the coolant and the pebbles beds technology that enable the on-line fueling and defueling. Since the FHR does not require lots of water as the coolant, it does not need a large amount of water. All components of FHR inside the size range that is rail shippable as well as road shippable, which is very important for the inland nuclear power plant construction. Consequently, FHR is a suitable option for construction of nuclear power plant in place where lack of water and has high dense population. With the help of comprehensive passive decay heat removal system, the FHR can remove the decay heat totally by natural circulation, without diesel generator as well as standby power supply grid. FHR employs Air-Brayton Combined Cycle (NACC) technology to run both on based load model and high peeking power load model, which can increase the income of nuclear power plant significantly. Furthermore, FHR employs many kinds of fuel cycle, especially for the full use of thorium resource. FHR can be used for hydrogen generation because the temperature of coolant is very high. As a result, the FHR related technology has become the research hotspot recently, which attracts the institutes and organizations from home and abroad to do research in this field. China will construct the world first FHR in 2020. However, since the current nuclear industry is highly dependent on water coolant technology, the fluoride salt related technology almost stop and most of the original data is derived from research in 1960s. In recent years, many originations from home and abroad start to revisit the fluoride salt related fundamental technology as well as experiment to design many kinds of fluoride-salt-cooled concept reactor system because the evolution of materials as well as the high requirement of the breakthrough of the bottleneck of the nuclear power development. The scope of this doctoral dissertation: Based on basic conservation law, a set of reasonable mathematical physical models has been built. The finite volume method, specifically semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations on staggered mesh is employed. Graph matrix theory is used to automatically generate a staggered mesh for large-scale complicated thermal hydraulics pipe network systems. Any position in the system can be set as reference point, where the static pressure and initial thermal properties start to propagate along system and complete the total static pressure distribution automatically as well as the check of close loop. The interpolation of interface of control volume is dependent on the kind of thermal information as well geometry characteristics, which is helpful to capture the tiny natural circulation mass flow rate. The numerical form of heat structure is based on fully implicit method, considering non-uniform grid, several layers of different materials and varying inner power distribution. Besides, heat structure model has three types of boundary conditions on each side. The combination of hydrodynamic model as well as heat structure model is flexible and can be coupled solved, which can be use for the prediction of peak temperature of heat structure as well the heat transfer rate and the parasitic heat loss. The reactor kinetics model, which consists of a set of stiff equations, is solved by backward differentiation formulae multi-step integration methods. Not only does this model can be solved faster, but also it can be coupled with hydrodynamic model as well as heat structure model by nuclear material Doppler effect as well as coolant, moderator temperature and density reactivity feedback model to predict the system response under varying conditions. Based on state of the art sparse matrix solver technology, the code can simulate characteristics of large-scale, multi-scale, complicated, multi-disconnected or connected, multi-branch thermal hydraulics pipe network system, including the mass flow distribution, pressure distribution, temperature distribution, energy transportation, natural circulation, heat transfer rate and parasitic heat loss of system. Control model that is based on comprehensive Boolean operation is used to simulate sequence of events for safety analysis of systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. As a result, the code has the capability to simulate many kinds of complicated safety accidents.
FHR employs non-uniform nuclear material with high power density, which is double- heterogeneous. The coolant is high Pr and viscosity temperature dependent at high temperature. The reactor core is porous media and the flow pattern changes dramatically with high heat transfer rate. The second flow in the coiled tubes make the pressure drop relatively higher but increases the heat transfer rate. The thermal properties of fluoride, nuclear material and other solid and liquid are summarized in this work. Based on fundamental analysis of fluoride heat transfer as well flow dynamic analysis, the closure models have been built including porous media reactor core, coiled tube air heater(CTAH), DRACS heat exchanger(DHX)，thermosyphon-cooled heat exchanger (TCHX), check valve, pipe component.
FHR advanced natural circulation analysis code (FANCY) for large-scale thermal hydraulics system analysis has been developed. The code has a user-friendly input deck and can run multi-input deck for sensitivity analysis. Under the Code Scaling and Analysis of Uncertainty(CSAU) framework, the verification and validation of code have been performed. The verification of code is performed by comparison of analytical solution with numerical solution. The sensitivity analysis of key parameter is also performed. Consequently, the accuracy of numerical model can be guaranteed, which increases the confidence of code prediction. The compact integral experiment test (CIET) is modeled, and then the validation of code is performed by comparison of simulation result with CIET natural circulation experimental data. The validation data include different boundary temperature of heat sink, varying power level, single loop natural circulation, coupled loop natural circulation. The accuracy of physical model in the code can be checked and uncertainty can by quantified.
The Mark 1 pebble-bed FHR system (MK1 PB-FHR) model with passive heat remove model and sequence of events model have been built. Given that power response affected by reactivity feedback, both the steady state as well as transient state can be simulated, including the normal flow and flow reversal, the transition from the normal force circulation to natural circulation. The safety analysis of FHR is based on the part of current identified safety license events of FHR: loss of offsite power and trip any pumps, loss of intermediate loop heat removal, the reactivity insertion from on-line fueling and defueling process as well as the change of porosity of pebble bed in normal operation. The failure close of DHX check valve to increase the bypass flow that reduce the coolant flow in the reactor. The sensitive analysis for the key parameters of the MK1 PB-FHR is performed, namely the influence of different reactivity insertion, the pump decay half time as well as the timing of the open check valve on the system response, which is important to quantify the uncertainty of safety analysis.
This work can be contributed to the design as well as licensing strategies for the fluoride-salt-cooled high temperature reactor technology(FHR)
Translated Keyword
[FHR thermal hydraulics passive heat remove system V&V safety analysis]
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