Yong-Bum Lee

CONCEPTUAL DESIGN OF THE SODIUM-COOLED FAST REACTOR KALIMER-600

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.

Model development for analysis of the Korea advanced liquid metal reactor

Abstract 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.

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.

SIMULATION OF THE EBR-II LOSS-OF-FLOW TESTS USING THE MARS CODE

Abstract 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 in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Modeling of flow blockage in a liquid metal-cooled reactor subassembly with a subchannel analysis code

Abstract The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal–cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the description of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.

DEVELOPMENT OF THE MATRA-LMR-FB FOR FLOW BLOCKAGE ANALYSIS IN A LMR

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5℃, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

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.

High-Temperature Design of Sodium-to-Air Heat Exchanger in Sodium Test Loop

KAERI가 개발 중인 소듐냉각 원형로의 붕괴열 제거를 위한 잔열제거계통(Decay heat Removal Circuit: DRC)은 안전등급으로 분류되며, 이는 설계의 다양성 확보를 위해 Fig. 1 에서와 같이 피동형 잔열제거계통(Passive DRC : PDRC)과 능동형 잔열제거계통(Active DRC : ADRC)으로 구성된다. 소듐대 공기 열교환기(sodium-to-air heat exchanger) 에는 ADRC 에 설치되는 핀형(finned) 소듐대 공기 열교환기인 FHX(Finned-tube Sodium-to-Air Heat Exchanger)와 PDRC에 설치되는 헬리컬형 소듐대 공기 열교환기인 AHX(Helical-coil Air Heat Exchanger)가 있다. 여기서는 이미Fig. 2 의 STELLA-1 소듐 시험루프 내에 설치된 AHX 와 KAERI 부지 내에 설치 예정인 Fig. 3 의 강제통풍형 소듐대 공기열교환기 Key Words: Sodium Test Loop(소듐 시험루프), Sodium-to-Air Heat Exchanger(소듐-공기열교환기), High Temperature Design(고온 설계), Creep-Fatigue(크리프-피로) 초록: 제4세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX 와 FHX 는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다. Abstract: In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

A new critical heat flux model for liquid metals under low heat flux-low flow conditions

Abstract The prediction of the critical heat flux (CHF) is an important consideration in the design and safety analysis of the sodium-cooled liquid metal fast breeder reactor (LMFBR). Many CHF correlations have been developed for light water reactor core applications. Compared with water, liquid metals show a divergent picture of boiling pattern. Therefore water CHF correlations cannot be applied directly to liquid metals. In this paper a mechanistic CHF model for liquid metals is developed based on the flow excursion mechanism. From the Baroczy correlation and the Ledinegg instability criterion a fundamental approach is tried to derive the relationship between CHF and the principal parameters. The overall mean accuracy ratio of the present model for 139 liquid metal CHF data points is 0.955, with a standard deviation of 0.155. Assessment shows that the predictions agree well with liquid metal CHF data within ±25% error bounds.

Theoretical investigation of vapor blanket velocity based on mass, energy, and momentum balance to predict CHF at low quality flow

Abstract In this paper a theoretical prediction method of the critical heat flux in flow boiling at low quality is presented based on the liquid sublayer dryout mechanism. The vapor blanket velocity which is the key parameter to predict the DNB type CHF is evaluated by solving mass, energy, and momentum balances simultaneously. The accuracy of the present model is evaluated by comparing model predictions with the experimental data for water. The predictions agree well with the extensive CHF data of water in uniformly heated vertical round tubes.