Translated Abstract
The Sodium-cooled fast reactor (SFR) is the most mature technology of the Generation IV advanced nuclear power systems, which has over 400 reactor years of operating experience. Due to the good capacity of breeding nuclear fuel and transmuting long-lived actinides, the SFR plays an important role in the strategy of sustainable development of nuclear energy. At present, China Demonstration Fast Reactor (CFR) is in the process of design and construction. Studies on thermal hydraulics and safety analysis of SFRs play an important role not only in offering reference for the design and safety review of CFR, but also in enhancing China international status in the field ofadvanced nuclear energy.
Based on the structure and operating characteristics of the pool-type SFRs, a series of physical and mathematic models were established. These models include component models for reactor core, inter-wrapper flow (IWF), pumps, passive decay heat removal systems, steam generators, heat exchangers, sodium pools, pipes, pipe network and relevant auxiliary models, etc. With these models, the non-uniform radial distribution of the core power and flow rate, the different core structures in axial direction, and some reactivity feedback effects are accurately simulated. The stratification phenomenon of the sodium pool under natural circulation condition can was simulated a the multi-grid pool model. A 2D layered IWF model was developed to evaluate the complicated thermal-hydraulic characteristics of IWF. The Transient Thermal-hydraulic Code for Analysis of Sodium Cooled Fast Reactor (THACS) was developed independently with the Fortran language. In this code, the Gear’s method was applied to solve the one-dimensional equations for the reactor system, and the SIMPLE algorithm was used for the solution of two-dimensional equations for IWF. Both parts were coupled with an explicit bi-directional format. The application of modular programming methods makes the THACS code more flexible.
Two benchmarks conducted by IAEA, namely the Shutdown Heat Removal Tests(SHRT) of EBR-II and the natural circulation test performed during the Phenix end of life experiments, were used to verify and test the THACS code. The protected loss of flow test (SHRT-17) and the unprotected loss of flow validate (SHRT-45R) of EBR-II were simulated with the THACS code. The good agreement between predicted results and measured data under natural circulation and forced circulation conditions confirms the accuracy and reasonability of these models in THACS code. The natural circulation test performed in Phenix was simulated with the THACS code, and the predicted results are agreed well with the measured data, which indicateds the application of THACS to different types of SFR and the accuracy and reliable of THACS was further verified. Moreover, study on the effect of IWF was conducted with the THACS code. The predicted results with IWF model are in a better agreement with the experiment data than that without IWF model, which proved the accuracy of IWF model. The comparison suggests that the IWF has a great influence on the thermal-hydraulic characteristics of the reactor core, and it can reduce the hot-spot temperature and flatten the temperature distribution in reactor core.
In this paper, the thermal-hydraulic behaviors and safety performance of Demonstration Sodium-Cooled Fast Reactor (DSFR) were studied with the THACS code. Six accidents were calculated, namely the reactivity insert accident (RIA), the loss of flow accident (LOFA), the loss of heat sink accident (LOHS), the unprotected transient overpower accident (UTOP), the unprotected loss heat sink accident(ULOHS), and the unprotected loss of flow accident (ULOF). The results show that DSFR can be safely shut down, and the coolant temperature and fuel temperature are within the safety limits under the protected accidents. Furthermore，the capability of three types of decay heat removal systems against a station blackout (SBO) accident was compared. These three decay heat removal systems are non-penetrating direct reactor auxiliary cooling system (NPDRACS), penetrating direct reactor auxiliary cooling system (PDRACS), and Primary Reactor Auxiliary Cooling System (PRACS), respectively. The results indicate that all the systems can remove the residual heat effectively. For large SFR, the capability of the penetrated DRACS and PRACS was better because the cold sodium from DHX or IHX could directly flow into the core assemblies. In the UTOP accident, a highest worth control rod was assumed to fully withdrawn from the core without control, DSFR operated at a new power level, and the peak temperatures of coolant, cladding and fuel were with safety limits. A sensitivity analysis of reactivity insertion was conducted, and the results show that the peak fuel temperature exceeded its melting point when the reactivity insertion was more than 620 pcm in 11 seconds. In the ULOHS accident, DSFR was safely shut down by the inherent negative reactivity feedback, and the residual heat in the core was removed effectively with the decay heat removal system. In the ULOF accident, the core temperatures increased rapidly because of the power-to-flow mismatch. The cladding temperature reached its temperature limit at 12 s, and the coolant began boiling at 20 s, and the core was in danger. The results show that with the application of the single failure criterion, DSFR could be shut down effectively in the ULOF accident, if three box of passive shutdown assemblies with a total reactivity worth -1.4% ΔK/K were in reactor core.
The research in this thesis can provide the tools the technical support to a pool-type sodium-cooled fast reactor in design and safety evaluation.
Translated Keyword
[Accident analysis, China Demonstration Fast Reactor, Inter-wrapper flow, Sodium-cooled fast reactor, THACS]
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