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In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical CO 2 Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

A heat transfer due to conduction through a coolant itself is not negligible in a liquid–metal cooled reactor (LMR). This portion of a heat transfer is frequently described with a conduction shape factor during the thermal-hydraulic design of an LMR. The conduction shape factor, which is highly dependent on a pitch-to-diameter ( P / D ) ratio, is defined as the ratio of the local conduction heat flux at a gap between two subchannels to the reference heat flux calculated by the averaged subchannel temperatures. The shape factors in heated triangular rod arrays for three different pitch-to-diameter ratios are generated through CFX calculations in the present study. The flow paths of 1.0–2.0 m in length are meshed into 180,000–360,000 volumes depending on the flow velocities. The SSG Reynolds stress model is used as a turbulent model in the calculations. The evaluated data fell between the heated-rod data and the plane-source data obtained by theoretical investigations. The conduction shape factors were found to be independent of the heating pattern of the rod arrays. Based on the evaluated data, a correlation for a liquid sodium coolant is suggested, which will improve the accuracy of the subchannel analysis codes for the thermal-hydraulic design of an LMR. When it is compared with the existing correlations, the suggested correlation is expected to enhance the reliability of the conduction shape factor because the data is evaluated by a more realistic numerical experiment.

The trends of turbulent mixing data in rod arrays and various definitions of the mixing factors are examined to derive a new definition of a mixing factor. The proposed mixing factor is based on the eddy diffusivity of energy and it takes into account the degree of the turbulent mixing of various fluids having different Prandtl numbers. With this definition of a mixing factor it is found that the geometrical parameter δ ij / D h is a dominant factor affecting a turbulent mixing, which correlates well with the reliable experimental data selected by considering the reliability of the measurement technique used in the experiments. A useful correlation of the turbulent mixing in rod arrays was developed as a function of δ ij / D h . The correlation is applicable for both square and triangular arrays with a reasonable accuracy. It predicts a reasonable mixing at a higher δ ij / D h or at a lower S / d ratio, which is the most improved feature of the correlation when compared with the existing ones. The proposed correlation may be applicable for the thermal-hydraulic design of a nuclear reactor with an improved accuracy.

The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.

In order to provide a consistent direction to long-term R&D activities, the Korea Atomic Energy Commission (KAEC) approved a long-term development plan for future nuclear reactor systems which include sodium cooled fast reactor (SFR) and very high temperature reactor (VHTR) on December 22, 2008. The SFR system is regarded as a promising technology to perform actinide management. The final goal of the long-term SFR development plan is the construction of an advanced SFR demonstration plant by 2028. The nuclear hydrogen project in Korea aims at designing and constructing a nuclear hydrogen demonstration system by 2022 to demonstrate its hydrogen production capability. This paper summarizes the overall long-term project plans for SFR and VHTR technology development and explains results of detailed design studies with supporting R&D activities.

The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced LIquid Metal Reactor), which is a sodium-cooled, metallic-fueled, pool-type reactor. The KALIMER-600 design concept (600 MWt) was selected as one of the reference GEN-IV sodium-cooled fast reactors (SFRs). The safety design philosophy of KALIMER-600 places maximum reliance on passive responses to abnormal and emergency conditions, and minimizes the need for active and engineered safety systems. KALIMER-600 utilizes the intrinsic negative reactivity feedback effect under design extended conditions where reactor scram failures are postulated. In order to assess the effectiveness of the inherent safety features, a scoping system analysis during unprotected overpower, loss of flow and under cooling events has been performed using the system-wide transient analysis code SSC-K. The present scoping analysis focuses on identification of enhanced safety design features that provide passive and self-regulating response to transient conditions and evaluation of safety margins. The results of the scoping analysis indicate an understanding of various inherent reactivity feedback mechanisms is very important in establishing design features. The analysis results show that the KALIMER-600 design concepts provide larger safety margins with respect to sodium boiling, fuel rod integrity, and structural integrity. The inherent safety can be enhanced by accounting for reactivity feedback mechanisms in the design process.Copyright © 2006 by ASME