No abstract available

[en] The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing an variety of off-normal or accident of a pool type design. It is developed at KAERI on the basis of SSC-L developed at BNL to analyze pool-type LMR transients. Because of inherent difference between th pool and loop design, the major modefications of SSC-L is required for the safety analysis of KALIMER. The major difference between KALIMER and general loop type LMRs exists in the primary heat transport system. In KALIMER, all of the essential components consisted of the primary heat transport system are located within the reactor vessel. This is contrast to the loop type LMRs, in which all the primary components are connected via piping to form loops attached externally to the reactor vessel. KALIMER has only one cover gas space. This eliminates the need for separate cover gas systems over liquid level in pump tanks and upper plenum. Since the sodium in hot pool is separated from cold pool by insulated barrier in KALIMER, The liquid level in hot pool is different from that in the cold pool mainly due to hydraulic losses and pump suction heads occuring during flow through the circulation pathes. In some accident conditions the liquid in the hot pool is flooded into cold pool and forms the natural circulation flow path. During the loss of heat sink transients, this will provided as a major heat rejection mechanism with the passive decay heat removal system. Since the pipes in the primary system exist only between pump discharge and core inlet plenum and are submerged in cold pool, a pipe rupture accident becomes less severe due to a constant back pressure exerted against the coolant flow from break. The intermediate and steam generator systems of both are generally identical. To adapt SSC-K to KALIMER design, the major modification of SSC-L has been made for the safety analysis of KALIMER. Test runs have been performed for the qualitative verification of the developed models. The present work would make it possible to use SSC-K for the priliminary safety analysis of KALMER. However, the further validation of SSC-K is required to be used for real applications. (Author). 5 refs., 3 tabs., 22 figs

No abstract available

[en] The design parameters of a commercial-scale fast breeder reactor which is currently under construction by regeneration of these data is preliminary analyzed. The analysis of nuclear and thermal characteristics as well as safety features of this reactor is emphasized. And the evaluation of the initial core mentioned in the system description is carried out in the areas of its kinetics and control system, and, at the same time, the flow distribution of sodium and temperature distribution of the initial FBR core system are calculated. (KAERI INIS Section)

[en] KALIMER is a pool type advanced liquid metal reactor which is being developed in KAERI. Advanced design features are incorporated into the conceptual design for the enhancement of its safety. However, for the ultimate safety of KALIMER, the containment dome design is introduced and analyzed. Before the performance analysis, The containment design of existing liquid metal reactor and the analysis methodology is reviewed. The methodology is established and test run for verification is performed. The codes for analysis are CONTAIN-LMR for containment thermal-hydraulic and aerosol behavior of containment, and MACCS for radiological consequence evaluation outside the containment. The preliminary containment dome design of KALIMER is determined to be single containment. The accidents for analysis are sodium pool fire and spray fire under HCDA condition. The source terms are determined for each accident and containment performance analyses have been performed. The sensitivity studies also have been performed with the containment dome volume. The results show that except 100 Kg sodium spray, the containment pressure is well established below the design limit. The containment dome is somewhat insensitive to containment volume. The exposure dose rates are estimated for each accidents, and show that the dose rate are below the PGA limit. The dose rate is more sensitive than containment pressure to containment volume. From the analysis results, the preliminary determined containment dome design is thought to ne rather optimal. Based on the analysis results, further sensitivity study on various parameters are required for design optimization. (Author). 8 refs., 15 tabs., 67 figs

[en] Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

[en] The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

No abstract available

[en] An effort has been made to analyze the hypothetical super-prompt-critical power excursions triggered by the reactivity insertion at an extremely large ramp rate into the sodium-voided core of a SFR. The system investigated in this study is the KALIMER-150 core, which is a pool-type sodium cooled prototype fast reactor that uses U-TRU-Zr metallic fuel to generate 392 MWt of power. The core utilizes a heterogeneous core configuration with the driver fuel and internal blanket zones alternately loaded in the radial direction. The density-dependent equations of state of pressure-energy density relationship developed for the metallic fuel U-Pu-Zr were implemented into the VENUS-II code, which was subsequently used in this study for analyzing the core disruptive accident. The equations of state for the single-phase liquid of the metallic uranium fuel were derived based on the van der Waals model, while a correlation recently developed for the saturated vapor pressure is used when the fuel is in the two-phase region. The results show that the extreme power excursion driven by inserting the reactivity at a rate of 100$/s is terminated without progressing into the single-phase region. (author)

[en] A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)