Yeong-Il Kim

Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR) design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

ADVANCED SFR DESIGN CONCEPTS AND R&D ACTIVITIES

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 semi-analytic multigroup nodal method

Abstract This paper describes a hybrid of the analytic nodal method and the polynomial nodal expansion method where the out-group source of the equivalent one-dimensional diffusion equation is approximated by a 4th-order polynomial. Then the higher-order source terms are updated in a weighted residual manner. The resulting equations are very similar to those of the nodal expansion method and are of a general multigroup form. The method has been demonstrated on several LWR benchmark problems, and the results show that the method gives the accuracy comparable to the analytic nodal method and is especially suitable for situations where steep flux gradients occur at assembly interfaces.

Computation of Turbulent Natural Convection in a Rectangular Cavity with the Lattice Boltzmann Method

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 in a rectangular cavity and the computed results are compared with the experimental data commonly used for the validation of turbulence models and those by the conventional finite-volume method. It is shown that the LBM with the present HYBRID thermal model predicts mean velocity components and turbulent quantities which are as good as those by the conventional finite-volume method. It is also found that the accuracy and stability of the solution is significantly affected by the treatment of the convection term, especially near the wall.

ACOUSTIC LEAK DETECTION TECHNOLOGY FOR WATER/STEAM SMALL LEAKS AND MICROLEAKS INTO SODIUM TO PROTECT AN SFR STEAM GENERATOR

Abstract 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 <1.0 g/s. The conditions and ranges for the existence of bubbling and jetting modes in water/steam outflow into circulating sodium through an injector device were determined to simulate a defect in the wall of the heat-transmitting tube of a sodium-water steam generator (SG). Based on experimental leak noise data, the simple dependency of the acoustic signal level on the leak rate of a microleak and small leaks at different frequency bands was presented for the principal analysis to develop an acoustic leak detection methodology for a KALIMER-600, 600-MW(thermal) reactor (K-600) SG, with the operational experiences for noise analysis and measurements of the Bystry neutron (fast neutron) reactor BN-600. Finally, the methodology was tested with the Korea Atomic Energy Research Institute (KAERI) acoustic leak detection system using sodium-water reaction signals of the Institute of Physics and Power Engineering and background noise of the Prototype Fast Reactor (PFR) superheater for methodology development of KAERI, and it was able to detect a leak rate of under 1 g/s and a signal-to-background noise ratio of −22 dB, using this system and methodology.

NUMERICAL ANALYSIS OF THERMAL STRATIFICATION IN THE UPPER PLENUM OF THE MONJU FAST REACTOR

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.

Status of Future Reactor Technology Development in Korea

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.Copyright

Pre-Test Analysis of Natural Circulation Test of PHENIX End-of-Life With the MARS-LMR Code

PHENIX, a prototype sodium-cooled fast reactor (SFR), has demonstrated a fast breeder reactor technology and also achieved its important role as an irradiation facility for innovative fuels and materials. In 2009 PHENIX reached its final shutdown and the CEA launched a PHENIX end-of-life (EOL) test program, which provided a unique opportunity to validate an SFR system analysis code. The Korea Atomic Energy Research Institute (KAERI) joined this program to evaluate the capability and limitation of the MARS-LMR code, which will be used as a basic tool for the design and analysis of future SFRs in Korea. For this purpose, pre-test analyses of PHENIX EOL natural circulation tests have been performed and one-dimensional thermal-hydraulic behaviors for these tests have been analyzed. The natural circulation test was initiated by the decrease of heat removal through steam generators (SGs). This resulted in the increase of intermediate heat exchanger (IHX) secondary inlet temperature, followed by a manual reactor scram and the decrease of secondary pump speed. After that, the primary flow rate was also controlled by the manual trip of three primary pumps. For the pre-test analysis the Phenix primary system and IHXs were nodalized into several volumes. Total 981 subassemblies in the core were modeled and they were divided into 7 flow channels. The active 4 IHXs were modeled independently to investigate the change of flow into each IHX. The cold pool was modeled by two axial nodes having 5 and 6 sub-volumes, respectively. The reactor vessel cooling system was modeled to match the flow balance in the primary system. The flow path of vessel cooling system was quite complicated. However, it is simplified in the modeling. For a MARS-LMR simulation, the dryout of SGs have been described by the use of the boundary conditions for IHTS as a form of time-to-temperature table. This boundary condition reflects the increase in IHTS temperature by SG dryout during the initial stage of the transient and the increase in heat removal by the opening of the two SG containments at 3 hours after the initiation of the transient. Through the comparison of the pre-analysis results with the prediction by other computer codes, it is found that the MARS-LMR code predicts natural circulation phenomena in a sodium system in a reasonable manner. The final analysis for validation of the code against the test data will be followed with an improved modeling in near future.Copyright

Evaluation of core nuclear analysis code system for LMR using measured physics parameters of BFS-73-1 critical assembly

Abstract As the first stage of the critical experiment plan for developing the Korea Advanced LIquid MEtal Reactor (KALIMER) core design, the basic neutronics characteristics of the BFS-73-1 critical assembly representing a uranium metal fueled benchmark core were investigated. The prediction capability of the core nuclear analysis code system, being developed as a standard analysis system in the KALIMER core design and analysis, has been evaluated against the BFS-73-1 experiment by comparing the calculated results to the measurements. The comparison shows that the effective multiplication factor was overpredicted by 0.3% in the C/E value. The 235 U and 238 U fission reaction rate distributions were underpredicted within a 2% discrepancy in the core region but appeared to be in poor agreement in the blanket region. The calculated values of spectral indices at the core center and the effective delayed neutron fraction, β eff , agreed with the measured ones within 2% and 1% discrepancies, respectively. The first order perturbation method predicted the sample reactivity worths for 235 U, 238 U, 239 Pu and 10 B within 7% but much larger discrepancies existed for the other materials. The calculated Doppler effect showed about an 8% overprediction.