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Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy. [ABSTRACT FROM AUTHOR]

• The Phenix natural circulation test was analyzed by system program THACS. • The predicted results of THACS are in good agreement with the test results. • The effect of Inter-wrapper flow on decay heat removal of pool-type SFR was analyzed. • The analytical capability of THACS program for natural circulation of pool-type SFR was evaluated. For the pool type sodium-cooled fast reactor (SFR), whether the primary circuit can establish a stable natural cycle after the reactor accident, and whether the primary natural cycle can take away the decay heat of the reactor core in the long-term, are the key issues in the design and safety analysis of SFR. In order to verify the natural circulation capacity of the primary circuit, a natural circulation test was carried out on pool type SFR Phenix by CEA before the reactor is decommissioned. One objective of this test is to verify the system codes used to simulate sodium natural convection phenomena. In this paper, the system code THACS is used to analyze Phenix natural circulation test, and the capacity of THACS to analyze the natural circulation of SFR is evaluated. Firstly, this paper introduces the components of Phenix reactor and the process of natural circulation test. Then, the models of program and the computational modeling of Phenix reactor are described. The calculation involves the whole process of the test: dry out of steam generators, scram, stop of pumps, two phases of natural circulation. Then, the results are compared with the test data, and the influence of inter-wrapper flow (IWF) on core outlet temperature is emphasized. The results show good applicability of THACS to analyze natural circulation in SFR, and the IWF can reduce the peak value of core outlet temperature at the beginning of natural cycle. Finally, the deviation between calculation and test results are analyzed, and the limitation of the models are evaluated. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [ABSTRACT FROM AUTHOR]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

• The PIRT of SBO for China pool type SFR has been developed by 15 experts. • A math method to evaluate different opinion about PIRT results has been proposed. • The results could be developed V&V matrix and a better understanding for SFR. In order to satisfy analysis of safety requirements for China pool type sodium-cooled fast reactor, a phenomena identification and ranking table (PIRT) is developed for station blackout (SBO) accidents by 15 experts. The SBO accidents could be divided into two phases, i.e. phase 1: force circulation and phase 2: natural circulation, and the most important phenomena but maximum disagreements among experts are the gap conductance model between fuel and clad, the inter wrapper flow, the heat transfer between coolant and inter wrapper. Those phenomena should be carefully assessed and done more theorical and experiment analysis to develop more accurate and realistic model, to reduce uncertainty of safety analysis. The PIRT could be also used for developing verification and validation matrix for thermal hydraulic safety code, guiding researchers and designers to conduct experiments to close technological gap and giving related researcher and designer a comprehensive understanding for pool-type sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]