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The thermal-hydraulic dynamics in containment are governed by a system of stiff ordinary differential equations (ODEs). A fully implicit discretization scheme is adopted to discretize these ODEs in order to mitigate the effects of stiffness. In comparison with explicit or semi-implicit discretization schemes that are subject to Courant limits on time steps, the fully implicit discretization scheme is more suitable for a containment analysis code that focuses on predicting both shortterm and long-term thermal-hydraulic parameters after an accident. This study introduces a general-purpose ODE solver for the containment analysis code. The outline of the solver is as follows: The fully implicit discrete equations lead to a large set of nonlinear equations that need to be solved using Newton's iterative method. The partial derivative components in the Jacobi matrix are calculated by the perturbation method using finite difference approximation, which avoids the complicated derivation of partial derivatives. The scaling modification technique is incorporated into this ODE solver to deal with significant differences in unknown variable magnitudes, and the line search method is introduced to address the difficulty of obtaining an accurate root estimate with Newton's method when the initial guess is far from the actual root. This proposed ODE solver was applied to two typical stiff ODE problems to test its stiffness-suppressed ability and to demonstrate that this proposed solver can perform calculations with a very large time step. Then, the CASSIA code, a containment analysis code developed by China Nuclear Power Technology Research Institute Co., Ltd (CNPRI), equipped with this ODE solver, was applied to the CSNI (Committee on the Safety of Nuclear Installations) benchmark problem and the Carolinas Virginia Tube Reactor (CVTR) test 3 problem to preliminarily demonstrate that the proposed ODE solver can perform containment thermal-hydraulic analysis correctly. This study could provide references for the development of a home-made containment analysis code. [ABSTRACT FROM AUTHOR]

Coupling supercritical carbon dioxide (S-CO2) Brayton cycle with Gen-IV reactor concepts could bring advantages of high compactness and efficiency. This study aims to design proper simple and recompression S-CO2 Brayton cycles working as the indirect cooling system for a mediate-temperature lead fast reactor and quantify the Brayton cycle performance with different heat rejection temperatures (from 32°C to 55°C) to investigate its potential use in different scenarios, like arid desert areas or areas with abundant water supply. High-efficiency S-CO2 Brayton cycle could offset the power conversion efficiency decrease caused by low core outlet temperature (which is 480°C in this study) and high compressor inlet temperature (which varies from 32°C to 55°C in this study). A thermodynamic analysis solver is developed to provide the analysis tool. The solver includes turbomachinery models for compressor and turbine and heat exchanger models for recuperator and precooler. The optimal design of simple Brayton cycle and recompression Brayton cycle for the lead fast reactor under water-cooled and dry-cooled conditions are carried out with consideration of recuperator temperature difference constraints and cycle efficiency. Optimal cycle efficiencies of 40.48% and 35.9% can be achieved for the recompression Brayton cycle and simple Brayton cycle under water-cooled condition. Optimal cycle efficiencies of 34.36% and 32.6% can be achieved for the recompression Brayton cycle and simple Brayton cycle under dry-cooled condition (compressor inlet temperature equals to 55°C). Increasing the dry cooling flow rate will be helpful to decrease the compressor inlet temperature. Every 5°C decrease in the compressor inlet temperature will bring 1.2% cycle efficiency increase for the recompression Brayton cycle and 0.7% cycle efficiency increase for the simple Brayton cycle. Helpful conclusions and advises are proposed for designing the Brayton cycle for mediate-temperature nuclear applications in this paper. [ABSTRACT FROM AUTHOR]

Supercritical CO2 Brayton cycle is a good choice of thermal-to-electric energy conversion system, which owns a high cycle efficiency and a compact cycle configuration. It can be used in many power-generation applications, such as nuclear power, concentrated solar thermal, fossil fuel boilers, and shipboard propulsion system. Transient analysis code for Supercritical CO2 Brayton cycle is a necessity in the areas of transient analyses, control strategy study, and accident analyses. In this paper, a transient analysis code SCTRAN/CO2 is developed for Supercritical CO2 Brayton Loop based on a homogenous model. Heat conduction model, point neutron power model (which is developed for nuclear power application), turbomachinery model for gas turbine, compressor and shaft model, and PCHE type recuperator model are all included in this transient analysis code. The initial verifications were performed for components and constitutive models like heat transfer model, friction model, and compressor model. The verification of integrated system transient was also conducted through making comparison with experiment data of SCO2EP of KAIST. The comparison results show that SCTRAN/CO2 owns the ability to simulate transient process for S-CO2 Brayton cycle. SCTRAN/CO2 will become an important tool for further study of Supercritical CO2 Bryton cycle-based nuclear reactor concepts. [ABSTRACT FROM AUTHOR]