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A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (∼300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

A numerical study of a turbulent natural convection in a rectangular cavity with the lattice Boltzmann method (LBM) is presented. The primary emphasis of the present study is placed on investigation of accuracy and numerical stability of the LBM for the turbulent natural-convection flow. A HYBRID method in which the thermal equation is solved by the conventional Reynolds-averaged Navier-Stokes equation (RANS) method while the conservation of mass and momentum equations are resolved by the LBM is employed in the present study. The elliptic-relaxation model is employed for the turbulence model and the turbulent heat fluxes are treated by the algebraic flux model. All the governing equations are discretized on a cell-centered, nonuniform grid using the finite-volume method. The convection terms are treated by a second-order central-difference scheme with the deferred correction method to ensure accuracy and stability of solutions. The present LBM is applied to the prediction of a turbulent natural convection...

This technical note presents the results of an experimental study of the role of water in sodium leak noise spectrum formation and at various water/steam leak rates of

An increase in the world population has accelerated the consumption of fossil fuels and deepened the pollution of global environment. As a result of these human activities, it is now difficult to clearly guarantee the sustainable future of humankind. An intuitional ‘must-go path’ scenario for the sustainable development of human civilization is proposed by extrapolating the human historical data over 30 years between 1970 and 2000. One of the most important parameters in order to realize the ‘must-go path’ scenario is the sustainability of energy without further pollution. In some countries an expanded use of nuclear energy is advantageous to increase sustainability, but fast reactor technology and closed fuel cycle have to be introduced to make it sustainable. In other countries, the development of cost-effective renewable energy, and the clean use of coal and oil are urgently needed to reduce pollution. The effect of fast nuclear reactor technology on sustainability as an option for near-term energy source is detailed in this paper. More cooperation between countries and worldwide collaboration coordinated by international organizations are essential to make the ‘must-go path’ scenario real in the upcoming 20 or 30 years.

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted...

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

The mechanical properties and microstructural evolution of modified 9Cr-1Mo steel have been studied to investigate steel property changes after long-term isothermal aging at 600 °C for 50,000 h. The microhardness and strength were maintained constantly after aging but the impact energy was dramatically reduced by 62 % during the aging period. From the viewpoint of microstructural evolution after the aging process, Cr-enrichment and Fe-depletion took place within the M23C6-type precipitates in the as-aged steel and V-depletion also happened within the VX-type precipitates after aging. In addition, the precipitates of the M2Mo-type Laves phase and the segregation of the impurity atoms would be formed during the long-term aging period. It was considered that the sharp reduction of the impact energy could be related to the formation of the Laves phases and the impurity segregation after aging at 600 °C. The phase stability was also verified by the specific heat results up to 950 °C from a DSC test. It was concluded from this study that the modified 9Cr-1Mo steel would keep its microstructural stability at 600 °C during the long-term aging period of 50,000 h, which was equivalent to the in-service life of the SFR fuel cladding.

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.

A local blockage in a subassembly of an LMR is of particular importance because the local temperature of the coolant increases at the downstream of a blockage and the integrity of the fuel clad can be threatened when an obstacle or a blockage is formed in a flow path. To analyze a flow blockage in a nuclear reactor core, Korea Atomic Energy Research Institute developed a subchannel analysis computer code MATRA-LMR-FB. This code adopts several enhanced modeling features such as a distributed resistance model, state-of-the-art turbulent mixing models, a hybrid difference scheme, and a porous body pressure drop model, therefore, it is applicable to a flow path with a plate-type or a porous-type blockage. The effect of each model has been evaluated through an analysis for the THORS experiment, in which a plate-type blockage was located in a flow channel with wire-wrapped fuel rods. The overall capability of the code has also been evaluated for the KNS experiment with a plate-type blockage in a grid-spaced flow channel, and for the SCARLET-II experiments with a porous blockage in channels formed by wire-wrapped fuel rods. The code shows good predictions for the experiments with a wire-wrapped flow path with a plate-type or a porous type blockage. The analyses for the KNS experiments reveal that the code requires a precise blockage model related to a grid spacer model.

Conformal mapped nodal simplified P 3 equations are derived and implemented for the two-dimensional neutronics analysis of fast reactor cores where hexagonal assemblies are loaded. The one-dimensional simplified P 3 equations are solved by using the nodal expansion method. The partial currents response matrices are constructed for the coupled simplified P 3 equations by applying the relationship between the partial currents and the surface-averaged fluxes. These matrices are then solved non-linearly. This method has been demonstrated for three two-dimensional fast reactor problems, and it was observed that the conformal mapped nodal simplified P 3 method can be effectively applied for the neutronics analysis of fast reactors with a substantial improvement in computational accuracy over the diffusion method.

The Super System Code of the Korea Atomic Energy Research Institute (SSC-K) has been developed for the transient analysis of the Korea Advanced Liquid MEtal Reactor (KALIMER) system. Recently, a detailed three-dimensional (3-D) core thermal-hydraulic model was developed to describe nonuniformities of radial temperature and flow within a subassembly and to decrease the uncertainties in the reactor safety margins during accident situations. The Shutdown Heat Removal Test-17 (SHRT-17) performed in the Experimental Breeder Reactor-II (EBR-II) and the postulated unscrammed events for the KALIMER conceptual design have been analyzed using a code system that has coupled a detailed 3-D core thermal-hydraulic model with SSC-K. The coupled code predicted behaviors for the experimental trends for the protected loss-of-flow SHRT-17. The KALIMER-150 design was adopted for a plant application of the same code system. Three events, unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS) were analyzed, and the simulation results were compared to those obtained using another code system that has coupled the Safety Analysis Section SYStem (SASSYS)-1 code with the same detailed 3-D core thermal-hydraulic model. The results, calculated with SSC-K coupled with the detailed 3-D core thermal-hydraulic model showed good agreement with the calculated results of the SASSYS-1 coupled code system for the UTOP and ULOF; however, some discrepancies were shown in the results for the ULOHS. These were found to have occurred because of a difference of the modeling for the decay heat removal system and primary coolant inventory. Through these analyses, the coupled code system was validated in order to be available for the safety analysis of a liquid-metal reactor (LMR) plant.

The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.

Based on a composition of 99.4 wt% AISI 316L stainless steel, 0.3 wt% Ti and 0.3 wt% Y2O3, an austenitic ODS steel was fabricated by a process of mechanical alloying, hot isostatic pressing and rolling. Fine oxide particles were observed in the matrix, and their chemical formulations were determined to be Y2Si2O7 and TiO. Heat treatment of the cold-rolled sample at 1200°C induced an isotropic tensile behavior at room temperature and at 700°C This result would be mainly attributed to the equiaxed grains that form as a result of the heat treatment for recrystallization.

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 a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

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

An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid Metal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examinedmore » the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations. (authors)« less