[en] Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

[en] In is essential to have an accurate prediction of core coolant and fuel temperature distribution in the liquid metal reactor (LMR) core themal hydraulic design, because of the design limits are imposed on the maximum temperatures of claddings and fuel pins in the sodium cooled LMRs. Due to the high thermal conductivity of the sodium, the transverse interassembly heat transfer may have a significant effect on the temperature profile within the subassembly, especially when it is adjacent to considerably hotter or colder subassemblies. Therefore, the interassembly heat transfer calculation should be considered in the LMR core thermal hydraulic design and analysis. For multi-assembly analysis, the interassembly heat transfe model was added in the MATRA-LMR code and the code was extended a single assembly analysis to multi-assembly analysis, i. e., a whole core code. For the assessment of the development status with interassembly heat transfer, the benchmark calculations were performed with SLTHE and THI3D codes using the 7-assembly problems. It is founded that the subassembly mixed mean coolant temperature has been changed as an effect of the interassembly heat transfer. And the maximum temperature change was found in the non-fueled subassembly which is considerabl colder than the fueled subassemblies

[en] Sodium cooled LMR core is comprised of many duct assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping module for an LMR core has been developed. This report describes the modeling and method of this module, and explains the calculation procedure and the sample calculation results. Firstly, LMR core thermal hydraulic conceptual design and analysis procedure was explained in chapter 1. Chapter 2 overviews this flow grouping module, and in chapter 3 core design and configuration data with power distributions were given. The calculation modeling and method of this module were explained in chapter 4, and chapter 5 shows calculation procedure and sample calculation results. KALIMER breeder core design data, e.g., inlet and outlet temperatures, power distributions and core flow, were used in this report to explain how this module works. And this module works in the environment of Microsoft Excel 2000 of MSOffice 2000

[en] Sodium cooled LMR core is generally comprised of many ducted assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping, assembly subchannel analysis and inter-assembly flow analysis have to be done in the LMR core thermal hydraulic design and analysis. This report describes this sodium cooled LMR core thermal hydraulic design procedure, in which are flow grouping, subchannel analysis and inter-assembly whole core analysis. And the French whole core analysis code system is described which is used for the domestic whole core thermal hydraulic analysis code system development. Firstly, sodium cooled LMR core thermal hydraulic conceptual design and analysis procedure is explained in chapter 2. Chapter 3 overviews the necessity and methodology of the whole core thermal hydraulic analysis, and the French whole core analysis system is described in chapter 4. Chapter 5 describes the domestic plan of the inter-assembly thermal hydraulic analysis system, and chapter 6 shows the conclusion and the future works

[en] Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the core temperature distribution to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR is being developed for LMR. The major modifications are as follows : A) The sodium properties table is implemented as subprogram in the code. B) Heat transfer coefficients are changed for LMR. C) The pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This user's guide describes code structure and equations of MATRA-LMR (Version 1.0), and explains input data preparation. (author). 19 refs., 7 tabs., 17 figs

[en] The main purpose of a liquid metal reactor core thermal-hydraulic design is to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power distribution in the core. The thermal-hydraulic design procedure consists of the coolant flow distribution to the subassemblies, the coolant/fuel temperature calculations and detailed subchannel analysis. But, there is a possibility to have a flow recirculation in the inter-assembly region because of the assembly flow resistance by the upper internal structure which is located above the core. It is necessary to analyze these phenomena to know the real temperature distribution in the assembly duct. There are only a few works in this area. This report describes the link calculation methodology and link codes, subchannel and porous medium approach codes, to perform the inter-assembly flow analysis, based on the LMR core thermal-hydraulic design methodology. And the preliminary calculation results for the inter-assembly flow calculation in the KALIMER-600 core were also described in this report

[en] Since the sodium boiling point is very high, maximum cladding and pin temperature are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the core temperature distribution to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR is being developed for LMR. The major modification are as follows : A) The sodium properties table is implemented as subprogram in the code. B) Heat transfer coefficients are changed for LMR C) The pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. To assess the development status of MATRA-LMR code, calculations have been performed for ORNL 19 pin and EBR-II 61 pin tests. MATRA-LMR calculation results are also compared with the results obtained by the ALTHEN code, which uses more simplied thermal hydraulic model. The MATRA-LMR predictions are found to agree well to the measured values. The differences in results between MATRA-LMR and SLTHEN have occurred because SLTHEN code uses the very simplied thermal-hydraulic model to reduce computing time. MATRA-LMR can be used only for single assembly analysis, but it is planned to extend for multi-assembly calculation. (author). 18 refs., 8 tabs., 14 figs

[en] This report, which summarizes the design concepts developed during Phase 4, follows the format of a safety analysis report. The purpose of publishing this report is to gather all of design information developed, so far in a systematic way, so that KALIMER-600 designers have a common and consistent source of for design information necessary for their future design and technology development activities on a SFR. Chapter 1 describes the KALIMER-600 Project. Chapter 2 includes the top-tier design requirements of KALIMER-600 and a general plant description. Chapter 3 summarizes the designs of the structures, components, equipment and systems. And the remaining chapters present the results of the design and safety analysis

[en] As a part of the core design development of KALIMER(150 MWe), the KALIMER core design which uses U-Zr binary fuel not in excess of 20% enrichment was performed. Starting from the former uranium metallic fueled core design, a more economic and safer equilibrium core design was first established based on extensive researches for the possible enrichment gains over various design options and in-core fuel management strategies. Further optimization to extend fuel discharge burnup has been achieved by employing strategic loading schemes for initial and transition cycles to reach the equilibrium cycle early. The core performance analysis based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and core average discharge burnup of 61.6 MWD/kg. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. When comparing with conventional plutonium metallic fueled cores of the same power level, the present KALIMER uranium metallic fueled core has an increased physical core size to meet the enrichment restriction, and, as a result, a lower power density to realize the minimum one-year cycle operation. The KALIMER uranium metallic fueled core characterized by its negative sodium void reactivity and low power density can be operated with maximizing its core safety characteristics as a first generation LMR. The present uranium metallic fueled core allows an easy replacement with different fuel compositions by its demands, with the accumulation of operation experience and design data verification. (author). 34 refs., 34 tabs., 12 figs.

[en] As a part of the core design development of KALIMER (150 MWe), the KALIMER core was initially designed with 20% enriched uranium metallic fuel. In this core design, the primary emphasis was given to realize the metallic fueled core design to meet the specific design requirements; 20% and below uranium enrichment and a minimum fuel cycle length of one year. The core was defined by a radially homogeneous core configuration incorporated with several passive design features to give inherent passive means of negative reactivity insertion. The core nuclear performance based on a once-through equilibrium fuel cycle scenario shows that the core has an average breeding ratio of 0.67 and maximum discharge burnup of 47.3 MWD/kg. When comparing with conventional plutonium metallic fueled cores of the same power level, the present uranium metallic fueled core has a lower power density due to its increased physical core size. The negative sodium void reactivity over the core shows a beneficial potential to assure inherent safety characteristics. The transition from the uranium startup to equilibrium cycle is feasible without any design change. Core nuclear performance characteristics in the present core design are attributed to the specific design requirements of enrichment restriction and fuel cycle length