Search

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

Thermal characteristics and mechanical analysis are particularly essential in the performance and safety analysis of metallic fuel for Liquid Metal Fast Reactor (LMFR). Therefore, the development of advanced thermo-mechanical analysis code is significant for the safety design of metallic fuels. In order to meet the future demand for metallic fuel application in China, the FRAC code was developed to analyze the metallic fuel performance during steady-state conditions. The code was benchmarked against the open-literature EBR-II experimental database. In comparison with the experimental data given, the code calculated Zr mole fraction, fuel temperature distribution, fission gas release and clad strain distribution are in good agreement with the data, which indicates that the code is reliable and can be applied into safety analysis of metal fuel design in China. The overall code assessment shows that the prediction error is within acceptable range. [ABSTRACT FROM AUTHOR]

Abstract: The concept of passive safety is put forward to improve the safety and reliability of the nuclear power plant. The secondary emergency passive residual heat removal system (EPRHRS) is an innovative design for China Pressurized Reactor (CPR1000), which can remove the core residual heat effectively and improve the inherent safety by passive methods when the station blackout (SBO) occurs. In this work, the EPRHRS, which is connected to the steam generator (SG) and consists of a heat exchanger (HX), a water tank and corresponding pipes and valves, was designed for CPR1000. The primary loop and EPRHRS were modeled using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system in the event of the SBO accident. In addition, sensitivity studies were also performed to investigate effects of the main parameters of the EPRHRS on the transient behavior of the primary loop and the EPRHRS. The effect of isolation valve opening time on the primary loop and the EPRHRS was also studied in the paper. The results of the study indicate that the optimized HX area is 200m2 when the EPRHRS water tank height is 10m and the optimized isolation valve opening time is 30s. The EPRHRS can remove the core decay heat from the primary loop effectively through the stable natural circulation established in the primary loop and the EPRHRS loop based on the calculation results. Therefore, it illustrates that the EPRHRS designed in the paper is effective and meaningful to CPR1000. [Copyright &y& Elsevier]

• 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]